[Federal Register Volume 67, Number 83 (Tuesday, April 30, 2002)]
[Proposed Rules]
[Pages 21389-21484]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 02-8108]



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Part III





Nuclear Regulatory Commission





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10 CFR Part 71



Compatibility With IAEA Transportation Safety Standards (TS-R-1) and 
Other Transportation Safety Amendments; Proposed Rule

Federal Register / Vol. 67, No. 83 / Tuesday, April 30, 2002 / 
Proposed Rules

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NUCLEAR REGULATORY COMMISSION

10 CFR Part 71

RIN 3150-AG71


Compatibility With IAEA Transportation Safety Standards (TS-R-1) 
and Other Transportation Safety Amendments

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is proposing to amend 
its regulations on packaging and transporting radioactive material to 
make them compatible with the International Atomic Energy Agency (IAEA) 
standards and to codify other applicable requirements. These changes 
would be compatible with ST-1 (TS-R-1), the latest revision of the IAEA 
transportation standards. This rulemaking would also address the 
unintended economic impact of NRC's emergency final rule entitled 
``Fissile Material Shipments and Exemptions'' (February 10, 1997; 62 FR 
5907) and a petition for rulemaking submitted by International Energy 
Consultants, Inc. (PRM-71-12: February 19, 1998; 63 FR 8362).

DATES: The comment period closes July 29, 2002. Comments received after 
this date will be considered if it is practical to do so, but the 
Commission is able to assure consideration only for comments received 
on or before this date.

ADDRESSES: Submit comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001. Attention: Rulemaking and 
Adjudications Staff.
    Deliver comments to 11555 Rockville Pike, Rockville, Maryland, 
between 7:30 a.m. and 4:15 p.m. on Federal workdays.
    You may also provide electronic comments via the NRC's interactive 
rulemaking website at http://ruleforum.llnl.gov. This site provides the 
capability to upload comments as files (any format), if your web 
browser supports that function. For information about the interactive 
rulemaking website, contact Ms. Carol Gallagher at (301) 415-5905 (e-
mail: [email protected]).
    Documents related to this action may be examined at the NRC Public 
Document Room (PDR) located at One White Flint North, 11555 Rockville 
Pike, Room O-1F23, Rockville, MD. Documents created or received at the 
NRC after November 1, 1999, are also available electronically at the 
NRC's Public Electronic Reading Room on the Internet at http://www.nrc.gov/reading-rm/adams.html. From this site, the public can gain 
entry into the NRC's Agencywide Documents Access and Management System 
(ADAMS), which provides text and image files of NRC's public documents. 
For more information, contact the NRC PDR Reference staff at 1-800-397-
4209, 301-415-4737, or email to [email protected].

FOR FURTHER INFORMATION CONTACT: Naiem S. Tanious, Office of Nuclear 
Material Safety and Safeguards, USNRC, Washington, DC 20555-0001, 
telephone: (301) 415-6103; e-mail; [email protected].

SUPPLEMENTARY INFORMATION:

Contents

I. Background
II. Summary of Public Comments
III. Request for Cost-Benefit and Exposure Information
IV. Discussion
    A. TS-R-1 Compatibility Issues
    Issue 1: Changing Part 71 to the International System of Units 
(SI) Only
    Issue 2: Radionuclide Exemption Values
    Issue 3: Revision of A1 and A2
    Issue 4: Uranium Hexafluoride Package Requirements
    Issue 5: Introduction of the Criticality Safety Index 
Requirements
    Issue 6: Type C Packages and Low Dispersible Material
    Issue 7: Deep Immersion Test
    Issue 8: Grandfathering Previously Approved Packages
    Issue 9: Changes to Various Definitions
    Issue 10: Crush Test for Fissile Material Package Design
    Issue 11: Fissile Material Package Design for Transport by 
Aircraft
    B. NRC-Initiated Issues
    Issue 12: Special Package Authorizations
    Issue 13: Expansion of Part 71 Quality Assurance Requirements to 
Certificate of Compliance (CoC) Holders
    Issue 14: Adoption of American Society of Mechanical Engineers 
(ASME) Code
    Issue 15: Change Authority for Dual-Purpose Package Certificate 
Holders
    Issue 16: Fissile Material Exemptions and General License 
Provisions
    Issue 17: Double Containment of Plutonium (PRM-71-12)
    Issue 18: Contamination Limits as Applied to Spent Fuel and 
High-Level Waste (HLW) Packages
    Issue 19: Modifications of Event Reporting Requirements
V. Section-By-Section Analysis
VI. Criminal Penalties
VII. Issues of Compatibility for Agreement States
VIII. Plain Language
IX. Voluntary Consensus Standards
X. Environmental Assessment: Finding of No Significant Impact
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII.Regulatory Flexibility Act Certification
XIV. Backfit Analysis

I. Background

    The Commission directed the NRC staff in Staff Requirements 
Memorandum (SRM) 00-0117 dated June 28, 2000: (1) To use an enhanced 
public-participation process (website and facilitated public meetings) 
to solicit public input on the part 71 rulemaking; and (2) to publish 
the staff's Part 71 issues paper in the Federal Register (65 FR 44360; 
July 17, 2000) for public comment. The issues paper presented the NRC's 
plan to revise Part 71 and provided a summary of all changes being 
considered, both IAEA-related changes and NRC-initiated changes. The 
NRC published the issues paper to begin an enhanced public-
participation process designed to solicit public input on the part 71 
rulemaking. This process included establishing an interactive website 
and holding three facilitated public meetings: a ``roundtable'' 
workshop at the NRC Headquarters, Rockville, MD, on August 10, 2000, 
and two ``townhall'' meetings--one in Atlanta, GA, on September 20, 
2000, and a second in Oakland, CA, on September 26, 2000.
    SRM-00-0117 also directed the staff to proceed, after completion of 
the public meetings, with the development of a proposed rule for 
submittal to the Commission by March 1, 2001. Oral and written comments 
received from the public meetings, by mail, and through the NRC 
website, in response to the issues paper, were considered in the 
drafting of the proposed changes contained herein.

Past NRC-IAEA Compatibility Revisions

    Recognizing that its international regulations for the safe 
transportation of radioactive material should be revised from time to 
time to reflect knowledge gained in scientific and technical advances 
and accumulated experience, IAEA invited Member States (the U.S. is a 
Member State) to submit comments and suggest changes to the regulations 
in 1969. As a result of this initiative, the IAEA issued revised 
regulations in 1973 (Regulations for the Safe Transport of Radioactive 
Material, 1973 edition, Safety Series No. 6). The IAEA also decided to 
periodically review its transportation regulations, at intervals of 
about 10 years, to ensure that the regulations are kept current. In 
1979, a review of IAEA's transportation regulations was initiated that 
resulted in the publication of revised regulations in 1985 (Regulations 
for the Safe Transport of Radioactive Material, 1985 edition, Safety 
Series No. 6).
    The NRC also periodically revises its regulations for the safe 
transportation of

[[Page 21391]]

radioactive material to make them compatible with those of the IAEA. On 
August 5, 1983 (48 FR 35600), the NRC published in the Federal Register 
a final revision to part 71, ``Packaging and Transportation of 
Radioactive Material.'' That revision, in combination with a parallel 
revision of the hazardous materials transportation regulations of the 
U.S. Department of Transportation (DOT), brought U.S. domestic 
transport regulations into general accord with the 1973 edition of IAEA 
transport regulations. The last revision to Part 71 was published on 
September 28, 1995 (60 FR 50248), to make part 71 compatible with the 
1985 IAEA Safety Series No. 6. The DOT published its corresponding 
revision to Title 49 on the same date (60 FR 50291).
    The last revision to the IAEA Safety Series 6 was named Safety 
Standards Series ST-1, published in December 1996, and was revised with 
minor editorial changes in June 2000, and was redesignated as TS-R-1. 
This rulemaking effort is to evaluate TS-R-1 for potential adoption in 
Part 71 regulations.
    Historically, the NRC coordinated its Part 71 revisions with DOT, 
because DOT is the U.S. Competent Authority for transportation of 
hazardous materials. ``Radioactive Materials'' is a subset of 
``Hazardous Materials'' in Title 49 regulations under DOT authority. 
Currently, DOT and NRC co-regulate transport of nuclear material in the 
United States. NRC is continuing with its coordinating effort with the 
DOT in this rulemaking process. Refer to the DOT's corresponding rule 
for additional background on the positions proposed in this notice.

Scope of 10 CFR Part 71 Rulemaking

    As directed by the Commission, NRC staff compared TS-R-1 to the 
previous version of Safety Series No. 6 to identify changes made in TS-
R-1, and then identified affected sections of Part 71. Based on this 
comparison, NRC staff identified 11 areas in part 71 that needed to be 
addressed in this rulemaking process as a result of the changes to the 
IAEA regulations. The staff grouped the part 71 IAEA compatibility 
changes into the following issues: (1) Changing part 71 to the 
International System of Units (Sl) (also known as the metric system) 
exclusively; (2) Radionuclide specific exemption values; (3) Revision 
of A1 and A2 values; (4) Uranium hexafluoride 
(UF6) package requirements; (5) Introduction of criticality 
safety index requirements; (6) Type C packages and low dispersible 
material; (7) Deep immersion test; (8) Grandfathering previously 
approved packages; (9) Adding and modifying Part 71 definitions; (10) 
Crush test for fissile material package design; and (11) Fissile 
material package design for transport by aircraft.
    Eight additional NRC-initiated issues (numbers 12 through 19) were 
identified by Commission direction, and through staff consideration, 
for incorporation in the Part 71 rulemaking process. These NRC-
initiated changes are: (12) Special package approvals; (13) Expansion 
of Part 71 quality assurance (QA) requirements to holders of, and 
applicants for, a Certificate of Compliance (CoC); (14) Adoption of the 
requirements of American Society of Mechanical Engineers (ASME), Boiler 
and Pressure Vessel (B&PV) Code for fabrication of spent fuel 
transportation packages; (15) Adoption of change authority; (16) 
Revisions to the fissile-exempt and general license provisions to 
address the unintended economic impact of the emergency rule (SRM-SECY-
99-200); (17) Decision on Petition for Rulemaking PRM-71-12, which 
requested deletion of the double containment requirements for 
plutonium; (18) Surface contamination limits as applied to spent fuel 
and high-level waste packages (SRM-SECY-00-0117); and (19) Part 71 
event reporting requirements. NRC published the first 18 issues in an 
issues paper in the Federal Register on July 17, 2000 (65 FR 44360).
    The Part 71 rulemaking is being coordinated with DOT to ensure that 
consistent regulatory standards are maintained between NRC and DOT 
radioactive material transportation regulations, and to ensure 
coordinated publication of the final rules by both agencies. On 
December 28, 1999 (64 FR 72633), DOT published an advance notice of 
proposed rulemaking regarding adoption of TS-R-1 in its regulations.

II. Summary of Public Comments

    The NRC held three public meetings to discuss and hear public 
comments on the issues under consideration for this rule. These 
meetings were transcribed by a court reporter; the meeting transcripts 
and condensed summaries of the comments made in the meeting are 
available to the public on the NRC's interactive rulemaking website at 
http:/ruleforum.llnl.gov and the Public Document Room located at One 
White Flint North, 11555 Rockville Pike, Room O-1F23, Rockville, MD. 
Also, the NRC received a total of 48 written comments on the issues 
paper during the meetings, by mail, and through the website. All of 
these written comments have been placed on the NRC website. The 
Commission has prepared a comment summary document entitled: ``Summary 
and Categorization of Public Comments on the Major Revision of 10 CFR 
Part 71.'' This document is published as NUREG/CR-6712, March 2001.
    This section provides a summary of general comments received at the 
public meetings that are not associated with any one issue, but rather 
with the NRC rulemaking process for this effort of the Part 71 
revision. A summary of public comments associated with a specific issue 
is included later in the discussion section under that issue. Comments 
not specific to this rulemaking effort are not included, nor are they 
discussed for their relevancy to the scope of this proposed action.

August 10, 2000, Meeting

    Two commenters supported moving towards risk-informed regulation 
because they believe it will increase the safety of nuclear power 
plants by allowing the operators to focus on risk-significant issues.
    Ten commenters wanted assurance that any changes to the NRC's 
regulations, whether in the context of conformity with international 
regulations, or solely affecting domestic shipments of radioactive 
materials, will not result in a reduction in transportation safety for 
the public.
    Two commenters suggested that NRC provide more information about 
the specific changes that will be incorporated into a proposed rule. 
One of these commenters also suggested that NRC consider increasing the 
number of public meetings and having them early on in the process in 
locations that will potentially be affected by any changes in the 
transportation regulations. The commenter also requested that the 
public comment period for this proposed rule be extended. This 
commenter also suggested that possibly by coordinating public meetings 
for all rulemakings or actions related to transportation (e.g., the 
Package Performance Study), the public will be better able to see the 
interrelation of the various NRC actions.
    Two commenters voiced their concern about the public accessibility 
of documentation related to transportation regulations. Specifically, 
they were concerned about the legal implications (i.e., due process) of 
not providing access to documents such as: (1) TS-R-1, (2) draft 
Advisory Material for the Regulations for the Safe Transport of 
Radioactive Material (TS-G-1.1) (supporting document for TS-R-1), and 
(3) the ASME code, while requesting public input on potential changes 
to the regulations to enhance conformity with

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international and domestic standards and regulations. One commenter 
noted that without these materials, the underlying basis of a proposed 
rule cannot be fully explored before its incorporation into the 
regulations.
    Two commenters were seeking clarification on the scope of the 
proposed changes. The commenters asked whether NRC intends to adopt all 
of the changes from IAEA's Safety Series 6 regulations that have been 
incorporated into the current TS-R-1 regulations, or just those 
identified in the proposed rule. One commenter also sought 
clarification as to whether the combined regulatory changes anticipated 
by NRC and DOT would cover all of the changes present in IAEA's TS-R-1 
regulations.
    Three commenters expressed concern over the possibility that the 
proposed changes in the transportation regulations could result in 
materials (including certain bulk materials) that were previously not 
regulated by NRC suddenly coming under NRC's jurisdiction, or actually 
becoming exempt in other jurisdictions. One commenter noted that this 
increased regulation could result in unnecessary concern on the part of 
the public as to the nature of the materials being transported. One 
commenter asked specifically if NRC was intending to start regulating 
naturally-occurring radioactive materials (NORM) and requested 
clarification on NRC's statutory authority to do so.
    One commenter suggested that, in addition to NRC and DOT, State 
agencies play an important role in the regulation of radioactive 
materials. The commenter noted that currently 32 States have entered 
into agreements with the NRC to become Agreement States. As Agreement 
States, they regulate use of radioactive material, and have regulations 
on transportation of radioactive material, including enforcement 
authority. The commenter is interested in being able to track possible 
changes in current regulations and how this could affect regulations at 
the State level.
    Seven commenters were concerned about the harmonization of NRC's 
regulations with those of the IAEA. The commenters expressed concern 
over the value of harmonization compared to the costs of 
implementation, and they further questioned the magnitude of the safety 
benefits of such harmonization. One commenter questioned that if Member 
States were not adopting TS-R-1 uniformly, what impact could that have 
on licensee's ability to transport internationally. Two commenters 
noted that while the TS-R-1 standards are burdensome, NRC does not want 
to stop commerce, and that is a risk if NRC does not adopt or harmonize 
with the TS-R-1 standards.
    Another commenter noted that the U.S. should have the right to 
adopt more stringent standards than those contained in TS-R-1. This 
commenter argued that uniform regulations should constitute a 
``minimum'' set of requirements and should not be considered the 
highest standard that should be applicable.
    One commenter suggested that NRC and DOT consider adopting a set of 
guiding principles to assure that harmonization is done in the best 
interest of public health and safety.
    Another commenter suggested that NRC adopt the IAEA regulations 
using a similar philosophy as is currently used by NRC, that is, by 
doing a safety check and ensuring that the level of safety is not 
diminished.
    Two commenters were seeking clarification on the authority of the 
international organizations over the activities of the U.S. The 
commenters suggested that if these organizations are directly 
influencing what U.S. regulatory agencies do, then the public has the 
right to more knowledge about their activities. One commenter suggested 
that any activity to harmonize international regulations with those of 
the U.S. should be done in open, accountable, democratic forums.

September 20, 2000, Meeting

    Several commenters were frustrated with the rulemaking process. 
These commenters indicated that a lack of easy access to pertinent 
resources, including TS-R-1 and relevant sections of the regulations, 
made it difficult to understand the nature, need, and potential impacts 
of the proposed changes. These commenters suggested that NRC seek 
alternative publication methods for relevant documents, such as posting 
the documents on the NRC website.
    Six commenters stated that NRC should only suggest changing 
existing standards if these changes improve or otherwise strengthen 
existing standards. Two commenters stated that attempting to affect any 
other change--i.e., not increasing the protection of public health and 
safety and the environment--is not worth its regulatory costs. However, 
if NRC is going to pursue these changes, then NRC should weigh heavily 
potential public and environmental costs. These commenters stated that 
while NRC is moving towards increased globalization, international 
standards should be considered a regulatory floor and not a ceiling. 
One commenter specifically cited that NRC should strengthen ``double-
casking requirements.''
    Three commenters stated that the proposed changes should not be 
allowed because they would increase public exposure rates without 
adequately informing the public of any risks associated with the 
increase. These commenters acknowledged the existence of background 
exposure rates, but believed that NRC needs to fully inform the public 
before changing current standards.
    Four commenters expressed an interest in better understanding the 
transportation process and the security arrangements associated with 
the proposed changes. One commenter specifically requested an 
explanation to what links existed between this rulemaking process and 
the NRC, the DOT, and the Department of Energy's (DOE's) currently 
scheduled shipments of radioactive materials. Another commenter 
requested an explanation on what security arrangements exist and what 
preparations NRC and DOT have made to deal with accidents and other 
such security breaches.
    One commenter suggested that the regulatory process be made as open 
and democratic as possible. This includes ensuring that supporting 
documents are not too expensive for the public to purchase, or 
otherwise access. Another commenter suggested that NRC hold additional 
public meetings to increase public involvement.

September 26, 2000, Meeting

    One commenter expressed his appreciation for the NRC using an 
enhanced rulemaking process and encouraged the NRC to continue using 
this process.
    Three commenters requested an extension of the public comment 
period to allow for additional public meetings. One commenter suggested 
that NRC hold not only additional public meetings, but also 
representative group sessions where Agreement States' representatives 
from affected cities, citizens' groups, and industry representatives 
discuss ``the substantive issues that are implicated by ST-1.''
    One commenter wanted to ensure that DOT and NRC have a process 
where NRC would jointly study and, after a reconciliation process, be 
able to address public comments in a coordinated fashion.
    Two commenters found it difficult to clearly identify what changes 
were being proposed. They requested additional details on the proposed 
changes and encouraged NRC to define all of the terms and provide 
background information in the next iteration.

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Specifically, they requested information that would enable the public 
to understand and evaluate the context and rationale for the proposed 
actions.
    Two commenters were concerned that NRC fully examine the impacts of 
the proposed changes on DOE as well as other Federal agencies, such as 
the U.S. Environmental Protection Agency (EPA). One of the commenters 
stated that, to date, he has not seen any such detailed analysis, an 
analysis the commenter requested at an earlier time. The commenter 
stated that when NRC has previously relaxed its standards, DOE has 
followed suit and cited the example of transportation standards.
    One commenter stated that NRC should view IAEA standards as 
minimum, not maximum, thresholds. The commenter requested that when 
NRC's regulations are more stringent than similar IAEA regulations, we 
retain that stringency. The commenter stated that he does not want NRC 
to lower its standards, and would prefer that international standards 
be raised.

Comments Received on the Website and by Mail

    Several commenters indicated the importance of adopting uniform 
regulations by all countries to ensure safe and uninterrupted 
transportation of radioactive materials internationally. The commenters 
indicated that the IAEA serves a vital role in developing regulations 
governing the international shipment of radioactive materials, and 
without this guidance, each country would develop its own regulations, 
thus making compatibility difficult, if not impossible, to achieve. 
These commenters strongly urged the NRC and DOT to make every effort to 
harmonize Part 71 with TS-R-1 regulations, as is reasonably achievable.
    Several commenters indicated that the public was not involved in 
the process that developed the TS-R-1 requirements. As a result, there 
is no objective analysis available for the public to determine which 
requirements are appropriate to change, and which ones are not.
    One commenter suggested that rather than NRC developing parallel 
regulations with DOT, NRC's regulations should only address those areas 
under NRC responsibility, such as fissile material and Type B 
shipments.
    Several commenters indicated that NRC must involve interested 
members of the public, State and local governments, and Tribes in a 
much broader framework in conjunction with the issuance of the proposed 
rule. One commenter argued that based on attendance at the public 
meetings, public participation has been inadequate and not 
representative. Another commenter noted that the public meetings were 
scheduled too close to the end of the public comment period, and that 
any meetings or hearings in conjunction with the proposed rule should 
be staged early in the comment process.
    One commenter suggested that the issues paper did not contain 
sufficient detail indicating the NRC's positions with respect to each 
of the issues. The commenter stated that inclusion of this information, 
including any regulatory drivers, would be helpful in furthering the 
public's understanding of the basis of these proposed changes, most 
specifically with respect to adoption of TS-R-1 requirements.
    One commenter raised the concern that the issues paper was not 
uniformly clear as to whether a proposed change would strengthen or 
weaken the protection of public health and safety in the U.S.
    One commenter was concerned that the proposal to harmonize NRC's 
regulations with international standards does not take into account the 
special nature of transportation in the U.S. For example, the commenter 
noted that a significant portion of the transportation occurs over 
distances exceeding 2,400 miles and often in rural areas, where 
emergency responders are volunteers with limited training. The 
commenter stated that regulations should be developed to protect 
emergency responders and other personnel who could be expected to be in 
contact with radioactive materials shipments.
    Several commenters requested an extension of the public comment 
period for the issues paper. The commenters cited several examples of 
why an extension is necessary, including impeded access to relevant 
information, periods of time during which the PDR was not open to the 
public, and closure of the Bibliographic Retrieval System for a period 
of 5 days.
    One commenter indicated that over the last several years, the 
majority of NRC rulemaking initiatives appear to be largely driven by 
concerns in providing regulatory relief for industry rather than in 
increasing safety for the public.
    One commenter claimed that IAEA standards are colored by 
consideration of commercial purposes. The commenter requested that NRC 
set aside commercial considerations in reviewing possible adoption of 
IAEA standards as NRC is first responsible to the American public and 
not to the international or domestic nuclear industry.
    Two commenters questioned whether NRC would take into account 
advances in science and engineering and accumulated experience since 
the development of the IAEA regulations 6 years ago. If not, one 
commenter argued that the proposed revisions to Part 71 could be 
outdated before they are issued.
    One commenter requested that TS-R-1 be made available for review to 
fully judge the impact that the proposed changes may have on 
transportation programs. For example, the commenter noted that one 
proposed change would result in different shipping names, without 
specifying those changes.
    One commenter suggested that NRC adopt a Transportation Safety Goal 
documenting the acceptable risk for the transportation of radioactive 
material.
    The public comments were considered in drafting the proposed 
requirements for 18 of the 19 issues (Issue 19 was added after 
publication of the issues paper). More details are provided under each 
issue.
    NRC has made copies of publicly released documents available on the 
website at http://www.nrc.gov/waste/spent-fuel-transp.html. 
Furthermore, the NRC plans to conduct additional public meetings during 
the proposed rule comment period. The dates and locations of these 
meetings will be noticed separately.

III. Request for Cost-Benefit and Exposure Information

    The NRC staff reviewed all public comments before drafting the 
proposed requirements in this notice. Summaries of all verbal, written, 
and electronic comments can be found in NUREG/CR-6712, Summary and 
Categorization of Public Comments on the Major Revision of 10 CFR Part 
71, March 2001). The staff also prepared a draft Regulatory Analysis 
(draft RA) to assess the economic impact of the proposed requirements. 
The draft RA is also published for public comment (for announcement, 
see Section XII).
    The NRC staff, as directed by the Commission, is continuing to 
solicit cost-benefit and exposure data from the public and industry to 
quantify the impact of the proposed Part 71 amendments. The NRC 
believes that this data will assist the Commission in: (1) Making an 
informed decision regarding the proposed IAEA compatibility changes, 
and (2) avoiding the promulgation of amendments that may result in 
unforeseen and unintended negative impacts, especially in view of the 
fact that the current regulations in Part 71 have provided adequate 
protection of the public health and safety.

[[Page 21394]]

    To help focus the public and industry and to capture the most data, 
the following request for information is presented in three groups: (1) 
General requests that apply to all 19 issues, (2) requests that apply 
only to the IAEA-related changes, and (3) issue-specific staff 
questions.

Request for Information on All 19 Issues

    The Commission is inviting comments from all stakeholders 
(Agreement States, public interest groups, and industry 
representatives) to address the overall impact of this proposed rule. 
Specifically, the Commission is soliciting: (1) Quantitative 
information and data on the costs and benefits which might occur if 
these proposed changes were adopted; (2) operational data on radiation 
exposures (increased or reduced) that might result from implementing 
the Part 71 proposed changes; (3) whether the proposed changes are 
adequate to protect public health and safety; (4) whether other changes 
should be considered, including providing cost-benefit and exposure 
data for these suggested changes; and (5) how should specific risk 
considerations (i.e., data on what can happen, how likely is it, what 
are the consequences) be factored into the proposed amendments.

Request for Information on the IAEA-Related Issues (Issues 1-11)

    The NRC recognizes the importance, from an international commerce 
standpoint, of having the packaging and transportation regulations in 
Part 71 compatible with the IAEA's TS-R-1. However, before adoption, 
the NRC seeks to quantify the impact of adopting these IAEA 
regulations. Development of the IAEA TS-R-1 did not directly involve 
the public or include a cost-benefit analysis. In contrast, NRC's 
practice is to consider costs and benefits in its regulatory analysis, 
and NRC is prepared to differ from the TS-R-1 standards, at least for 
domestic purposes, to the extent the standards cannot be justified from 
a cost-benefit perspective, especially given the current regulations in 
Part 71 have provided adequate protection of the public health and 
safety.
    Therefore, the NRC is inviting public comments on the IAEA-related 
issues, Issues 1-11. Specifically, the Commission is soliciting cost-
benefit data to quantify the economic impact of harmonizing with the 11 
IAEA changes on the domestic commerce and international commerce of 
packages containing radioactive material. The NRC is interested in 
determining: (1) whether the benefits of harmonization with the IAEA 
standards may exceed the costs, or may result in other health and 
safety problems resulting from dual standards between domestic (Part 
71) and international (TS-R-1) requirements, and (2) whether the NRC 
should adopt only some of the 11 IAEA changes.

Request for Responses to Issue-Specific Questions:

Issue 2--Radionuclide Exemption Values
    What impacts, if any, would result for industries that possess, 
use, or transport materials currently exempt from regulatory control 
(e.g., unimportant source material under 10 CFR 40.13) if adoption of 
the radionuclide exemption values were to occur in Part 71?
    What impacts, if any, would result for industries that transport 
natural material and ores containing naturally-occurring radionuclides 
which are not intended for processing for economic use of their 
isotopes (e.g., phosphate mining, waste products from the oil and gas 
industry), if the TS-R-1 exemption values are adopted, but without the 
``10 times the applicable exemption values'' provision?
    Another possible impact of the proposed radionuclide exemption 
values is in the area of waste disposal sites which are regulated by 
EPA under the Resource Conservation and Recovery Act (RCRA). The 
acceptance limit in these sites for materials containing radioactive 
residuals is the existing 70 Bq/g (0.002 Ci/g) standard used 
by DOT, NRC, and EPA. Presently, only the NRC and DOT are proposing to 
adopt the exemption values, which may result in situations where 
shipment of materials with residual radioactivity would be allowed for 
transportation under the new exemption values but would not be allowed 
for disposal in RCRA sites.
    What cost impacts or other problems, if any, would result from 
adoption of the exemption values, in Part 71 and DOT regulations, for 
industries or entities involved in the shipment and disposal of 
materials with residual activity to RCRA sites?
Issue 3--Revision of A1 and A2
    What impacts, if any, would result for the radiopharmaceutical 
industry in terms of cost and worker dose by adopting the lower 
international A2 value, rather than retaining the current 
A2 value for domestic shipment of molybdenum-99?
    What impacts, if any, would result for industry in terms of cost 
and worker dose by retaining the current A1 and 
A2 values for californium-252, rather than adopting the 
international A1 and A2 values?
    What impacts, if any, would result for industry in terms of cost 
and worker dose by not including in Table A-1 (A1 and 
A2 Values for Radionuclides) the 16 radionuclides that are 
listed in the current Part 71 but not in TS-R-1?
Issue 4--Uranium Hexafluoride UF6 Package Requirements
    Should the current practice of excluding moderators in criticality 
evaluations for UF6 packages be continued?
Issue 5--Introduction of the Criticality Safety Index Requirements
    What cost or benefit impacts would result if the per package 
Criticality Safety Index (CSI) were to change from 10 to 50?
Issue 6--Type C Packages and Low Dispersible Material
    NRC requests information on the need for Type C packages, 
specifically on the number of package designs and the timing of future 
requests for Type C package design approvals.
Issue 8--Grandfathering Previously Approved Packages
    Under what conditions should packagings be removed from service?
    What are the cost or benefit impacts associated with the proposal 
to remove B( ) packages from service?
Issue 10--Crush Test for Fissile Material Package Design
    What are the cost or benefit impacts of imposing the crush test 
requirement on fissile material package designs?
Issue 12--Special Package Approval
    What additional limitations, if any, should apply to the conditions 
under which an applicant could apply for a package authorization?
Issue 17--Double Containment of Plutonium (PRM-71-12)
    What cost or benefit impacts would arise from removal of the double 
containment requirement for plutonium?
Issue 18--Contamination Limits as Applied to Spent Fuel and High-Level 
Waste (HLW) Packages
    NRC requests information regarding the application of the 
regulatory limits for removable contamination on the external surfaces 
of packages used for spent fuel shipments. This information will be 
most helpful if respondents also

[[Page 21395]]

indicate the cask design used and whether or not the cask is fitted 
with a protective cover prior to immersion in the spent fuel pool. 
Specifically, for previous spent fuel shipments, information is sought 
on:
    (1) The removable contamination level on the cask surface after the 
cask has been loaded, removed from the spent fuel pool, and dried;
    (2) The dose attributable to any decontamination efforts, including 
external dose from cask and facility radiation fields and internal dose 
from airborne radioactivity in the cask handling/loading areas;
    (3) The removable contamination level on the cask surface after 
decontamination efforts and before shipment; and
    (4) The removable contamination levels on the cask surface upon 
receipt at the destination facility.

IV. Discussion

    This section is structured to present and discuss each issue 
separately (with cross references as appropriate). Each issue has four 
parts: Background, Discussion, NRC Proposed Position, and Affected 
Sections. The discussion section summarizes the public comments, NRC 
staff consideration of public comments and of technical and policy 
issues, and the regulatory analysis for that issue.

A. TS-R-1 Compatibility Issues

Issue 1. Changing Part 71 to the International System of Units (SI) 
Only
    Background. TS-R-1 uses the SI units exclusively. This change is 
stated in TS-R-1, Annex II, page 199: ``This edition of the Regulations 
for the Safe Transport of Radioactive Material uses the International 
System of Units (SI).'' The change to SI units exclusively is evident 
throughout TS-R-1. TS-R-1 also requires that activity values entered on 
shipping papers and displayed on package labels be expressed only in SI 
units (paragraphs 543 and 549). Safety Series No. 6 (TS-R-1's 
predecessor) used SI units as the primary controlling units, with 
subsidiary units in parentheses (Safety Series 6, Appendix II, page 
97), and either units were permissible on labels and shipping papers 
(paragraphs 442 and 447).
    The TS-R-1 change is in conflict with the NRC Metrication Policy 
issued on June 19, 1996 (61 FR 31169), which allows a dual-unit system 
to be used (SI units with customary units in parentheses). The NRC 
Metrication Policy was designed to allow market forces to determine the 
extent and timing for the use of the metric system of measurements. The 
NRC is committed, in that policy, to work with licensees and applicants 
and with national, international, professional, and industry standards-
setting bodies [e.g., American National Standards Institute (ANSI), 
American Society for Testing and Materials (ASTM), ASME] to ensure 
metric-compatible regulations and regulatory guidance. The NRC 
encouraged its licensees and applicants, through its Metrication 
Policy, to employ the metric system wherever and whenever its use is 
not potentially detrimental to public health and safety, or its use is 
economic. The NRC did not make metrication mandatory by rulemaking 
because no corresponding improvement in public health and safety would 
result, but rather, costs would be incurred without benefit. As a 
result, licensees and applicants use both metric and customary units of 
measurement.
    According to the NRC's Metrication Policy, the following documents 
should be published in dual units (beginning January 7, 1993): new 
regulations, major amendments to existing regulations, regulatory 
guides, NUREG-series documents, policy statements, information notices, 
generic letters, bulletins, and all written communications directed to 
the public. Documents specific to a licensee, such as inspection 
reports and docketed material dealing with a particular licensee, will 
be issued in the system of units employed by the licensee.
    Currently, Part 71 uses the dual-unit system in accordance with the 
NRC Metrication Policy.
    Discussion. Oral comments received at the public meetings, as well 
as written comments received on the issues paper, indicate opposition 
to the use of SI units only. Most commenters were opposed to switching 
to SI units only, and supported the continued use of the dual-unit 
system. At the August 10 meeting, a radiopharmaceutical industry 
representative commented that the Food and Drug Administration (FDA) 
requires the use of customary units (curie units), while shipping 
papers always list the activity in becquerels with curies in 
parentheses. The representative stated that while that presents some 
problems now, the industry is able to handle it. By moving to a system 
where the shipping papers are in SI units only, a situation would be 
created where the package contents are expressed in curies, while 
shipping papers and labels are expressed in becquerels. This could be 
confusing, especially when comparing the shipping papers to the 
contents. The implication is that this situation could create 
complications at the shipment destination as personnel would have to 
perform unit conversions to match package contents with the shipping 
papers. Furthermore, there was a concern that this could result in 
errors in patient administrations. Other commenters indicated that this 
change would result in significant costs for industry, with no apparent 
safety benefit.
    Another commenter indicated that, although the U.S. has adopted a 
policy of shifting to SI units, this policy has not been implemented. 
Several commenters argued that requiring the use of SI units only for 
domestic shipments of radioactive materials, when the balance of the 
nation's activities are conducted in customary units, would cause 
confusion as well as possible safety issues if misunderstandings or 
miscalculations were to occur. The commenters noted that the majority 
of individuals (including emergency response workers) are more 
accustomed to using customary units, and by requiring the use of SI 
units, problems would occur in converting customary units to SI units. 
As a result, the commenters believed that this could result in an 
increased risk of inadvertent exposure of workers to radiation.
    One commenter indicated that SI units are currently required to be 
used in certain cases for shipping and believed that such a change 
would pose little risk. However, the commenter added that any such 
change should be accompanied by a 3-year delay in the effective date to 
allow for proper transition.
    NRC staff notes that the use of SI units only would conflict with 
the NRC's Metrication Policy, which allows the use of a dual-unit 
system for measurements. The statement made in NRC's final Metrication 
Policy, ``* * * the NRC believed and continues to believe that if 
metrication were made mandatory by a rulemaking, no corresponding 
improvement in public health and safety would result but costs would be 
incurred without benefit,'' still stands.
    The NRC draft regulatory analysis (draft RA) indicates that 
maintaining the existing policy of allowing the use of dual units is 
appropriate from a safety, regulatory, and cost perspective. A change 
to require SI units only would necessitate an exemption by the 
Commission from its dual-units policy, and would result in an 
inconsistency between Part 71 and other parts of the Commission's 
regulations. Further, anticipated costs to industry for implementing 
the new requirement

[[Page 21396]]

(e.g., training, recalculations), estimated to be between $12.6 and 
$16.3 million, would be avoided if the dual-unit system is maintained. 
In addition, while NRC would incur $15,000 in costs by converting from 
one system of units to another, this cost is offset by a savings in 
resources for not proceeding with rulemaking activities to implement 
the change. As discussed by several commenters, the change to SI units 
only could result in the potential for adverse impact on the health and 
safety of workers and the general public as a result of unintended 
exposure in the event of shipping accidents, or medical dose errors, 
caused by confusion or erroneous conversion between the currently 
prevailing customary units and the new SI units by emergency responders 
or medical personnel.
    The NRC considered the Commission policy on this issue, the above 
public comments, and the draft RA of the impact of this change, and 
concluded that adopting the IAEA use of SI units only in Part 71 would 
have both a cost impact and potentially negative impact on workers and 
public health and safety.
    NRC Proposed Position. The NRC does not intend to change Part 71 to 
use SI units only, nor does it intend to impose on Part 71 licensees, 
certificate holders, or applicants for a CoC the use of SI units only. 
While TS-R-1 uses SI units only, it does not specifically prohibit the 
use of a dual-unit system (SI units and customary units). Therefore, 
the NRC will continue to use the dual-unit system in Part 71.
    Affected Sections. None (not adopted).
Issue 2. Radionuclide Exemption Values
    Background. The DOT currently uses a specific activity threshold of 
70 Bq/g (0.002 Ci/g) for defining a material as radioactive 
for transportation purposes. DOT regulations apply to all materials 
with specific activities that exceed this value. Materials are exempt 
from DOT's transportation regulations if the specific activity is equal 
to or below this value. The 70-Bq/g (0.002-Ci/g) specific 
activity value is applied collectively for all radionuclides present in 
a material.
    Within Sec. 71.10, the NRC uses the same specific activity 
threshold as a means of determining if a radioactive material is 
subject to the requirements of Part 71. Materials are exempt from the 
transportation requirements in Part 71 if the specific activity is 
equal to or below this value. Although the materials may be exempt from 
any additional transportation requirements under Part 71, the 
requirements for controlling the possession, use, and transfer of 
materials under Parts 30, 40, and 70 continue to apply, as appropriate, 
to the type, form, and quantity of material.
    During the development of TS-R-1, it was recognized that there was 
no technical justification for the use of a single activity-based 
exemption 70-Bq/g (0.002-Ci/g) value for all radionuclides. It 
was concluded that a more rigorous technical approach would be to base 
radionuclide exemptions on a uniform dose basis, rather than a uniform 
specific activity (also known as activity concentration) basis.
    By 1994, the IAEA and other international health-related 
organizations had developed the International Basic Safety Standards 
for Protection against Ionizing Radiation and for the Safety of 
Radiation Sources, IAEA Safety Series No. 115. (This document is 
sometimes referred to informally as the Basic Safety Standards, or 
BSS.) During the preparation of this document, a set of principles had 
been developed and accepted for determining when exemption from 
regulation was appropriate. One of the exemption criteria was that the 
effective dose expected to be incurred by a member of the public from a 
practice (e.g., medical use of radiopharmaceuticals in nuclear medicine 
applications) or a source within a practice should be unlikely to 
exceed a value of 10 Sv (1 mrem) per year. IAEA Member State 
researchers developed a set of exposure scenarios and pathways which 
could result in exposure to workers and members of the public. These 
scenarios and pathways were used to calculate radionuclide exemption 
activity concentrations and exemption activities which would not exceed 
the recommended dose (see Safety Series No. 115, Schedule I, 
``Exemptions'').
    To investigate the exemption issue from a transportation 
perspective during the development of TS-R-1, IAEA Member State 
researchers calculated the activity concentration and activity for each 
radionuclide that would result in a dose of 10 Sv (1 mrem) per 
year to transport workers under various BSS and transportation-specific 
scenarios. Due to differences in radionuclide radiation emissions, 
exposure pathways, etc., the resulting radionuclide-specific activity 
concentrations varied widely. The appropriate activity concentrations 
for some radionuclides were determined to be less than 70 Bq/g (0.002 
Ci/g), while the activity concentrations for others were much 
greater. However, the calculated dose to transport workers that would 
result from repetitive transport of each radionuclide at its exempt 
activity concentration was the same [(10 Sv) (1 mrem)] per 
year. For the single activity-based value, the opposite was true, i.e., 
the exempt activity concentration was the same for all radionuclides 
(70 Bq/g) (0.002 Ci/g), but the resulting doses under the same 
transportation scenarios varied widely, with annual doses ranging from 
much less than 10 Sv (1 mrem) per year for some radionuclides 
to greater than 10 Sv (1 mrem) per year for others. The 
radionuclide-specific activity concentration values reduced the 
variability in doses that were likely to result from exempt transport 
activities.
    IAEA noted that the exempt activity concentrations calculated for 
transportation scenarios were less than those found in Safety Series 
No. 115 (BSS), Table I-I, ``EXEMPTION LEVELS: EXEMPT ACTIVITY 
CONCENTRATIONS AND EXEMPT ACTIVITIES OF RADIONUCLIDES (ROUNDED)'', but 
not by more than a factor of 100. IAEA did not believe the differences 
warranted a second set of exemption values, and therefore adopted the 
Safety Series No. 115 (BSS) values in TS-R-1. These values are found in 
TS-R-1, paragraphs 401-406, and in Tables I and II.
    A consequence of using the BSS exemption values for transportation 
is that the estimated average annual dose under the transportation 
scenarios exceeds the 10 Sv (1 mrem) per year criterion for 
some radionuclides. The staff has estimated that the average annual 
dose per radionuclide under the transportation scenarios using the BSS 
exemption values for a representative list of 20 radionuclides is 0.25 
mSv (25 mrem) per year. However, the staff estimates that the 
corresponding dose for the current 70 Bq/g (0.002 Ci/g) 
exemption value, using the same transportation scenarios and 
radionuclides, is approximately 0.5 mSv (50 mrem) per year. Although 
both the current exemption value and the BSS exemption values result in 
an estimated average dose per radionuclide that exceeds the criterion, 
the dose estimated for the BSS exemption values is significantly less 
than that estimated for the current 70 Bq/g (0.002 Ci/g) 
exemption value.
    Note that some nuclides listed in Table I have a reference to 
footnote (b). These nuclides have the radiological contributions from 
their daughter products (progeny) already included in the listed value. 
For example, natural uranium [U (nat)] in Table I has a listed activity 
concentration for exempt material of 1 Bq/g (2.7 x 10-5 Ci/g). 
This means the activity concentration of the uranium is limited to 1 
Bq/g (2.7 x 10-5 Ci/g), but the total activity

[[Page 21397]]

concentration of an exempt material containing 1 Bq/g (2.7 x 10-5 
Ci/g) of uranium will be higher (approximately 7 Bq/g (1.9 x 
10-4 Ci/g)) due to the radioactivity of the daughter products.
    The basis for the exemption values, as discussed in the draft 
Advisory Material for the Regulations for the Safe Transport of 
Radioactive Material, TS-G-1.1, paragraphs 107.5 and 401.3, indicates 
that materials with very low hazards can be safely exempted from the 
transportation regulations. If the exemptions did not exist, enormous 
amounts of material with only slight radiological risks, materials 
which are not ordinarily considered to be radioactive, would be 
unnecessarily regulated during transport.
    Based on TS-R-1, paragraph 236, when both the activity 
concentration for exempt material and the activity limit for an exempt 
consignment are exceeded, the material or consignment must meet 
applicable transportation regulations. Paragraph 404 of TS-R-1 
specifies how exemption values may be determined for mixtures of 
radionuclides.
    Some of the lower activity concentration values might include NORM. 
As an example, ores may contain NORM. In regard to transporting NORM, 
one petroleum industry representative stated there are no findings that 
indicate the current standard fails to protect the public, and that 
there is no benefit in making the threshold more stringent. Further, it 
would have a significant impact on their operations. Other similar 
comments were received during the public meetings. The overall impact 
would be that some material formerly not subject to the radioactive 
material transport regulations may need to be transported as 
radioactive material and therefore meet the corresponding applicable 
DOT transport requirements.
    IAEA recognized that application of the activity concentration 
exemption values to natural materials and ores might result in 
unnecessary regulation of these shipments, and established a further 
exemption for certain types of these materials. Paragraph 107(e) of TS-
R-1 further exempts: ``natural material and ores containing naturally 
occurring radionuclides which are not intended to be processed for use 
of these radionuclides provided the activity concentration of the 
material does not exceed 10 times the values specified in paragraphs 
401-406.''
    Discussion. Comments were received on this issue during the public 
meetings, by mail, and on the NRC web site. One commenter stated that 
the NRC should reference all DOT equivalent regulations (the 
radionuclide exemption values and all others) to prevent conflict 
between the NRC and DOT regulations. Two commenters cautioned that 
moving from one exemption value to different values for each 
radionuclide could result in more complicated compliance and 
enforcement scenarios. For example, one commenter indicated that the 
70-Bq/g (0.002-Ci/g) exemption limit is also used as a 
standard by EPA under the RCRA as the permit limit for the acceptance 
of material containing radioactive residuals. Any changes to this limit 
could result in the preclusion of certain materials for disposal at 
permitted disposal facilities. Some commenters indicated that the 
revised exemption values should apply not only to domestic shipments 
but to exported shipments as well.
    One commenter indicated that this change will have a significant 
unintended impact on its operations because most of the oil and gas 
shipments would not be exempt under the new rule.
    One commenter indicated that such a change would result in an 
increase in the number of shipments by requiring smaller quantities to 
be shipped due to the lower exemption values. Another commenter 
suggested that the use of radionuclide-specific exemption values would 
not result in an increase in the number of packages being shipped, but 
would result in more shipments being labeled as radioactive. The 
commenter argued that because many of these shipments are currently 
being made as ``nonhazardous'' shipments, many of the responses to 
accidents will be for minimal hazard materials representing 
insignificant risks that do not warrant increased response safety. The 
commenter stated that this would not result in increased safety, but 
would instead divert emergency response personnel from other, more 
significant, tasks.
    Several commenters reflected a belief that, for some radionuclides, 
the new higher values would be a relaxation of the regulations, and 
thus will adversely impact public health and safety. A few commenters 
indicated that NRC should actually look at making the exemption values 
more stringent rather than reducing the level of protection currently 
afforded the public. One commenter suggested that, before adopting any 
of the exemption values contained in TS-R-1, NRC should scrutinize the 
values to determine whether they are justified as protective of human 
health and the environment.
    A few commenters supporting the retention of the current Part 71 
exemption values indicated that a move to radionuclide-specific 
exemption values would result in increased costs while yielding no 
additional safety benefit.
    The overall impact would be that some previously exempted material 
may need to be transported as radioactive material and therefore would 
need to meet applicable DOT transport requirements. While these 
activity concentration values would impact certain sectors, the NRC 
staff believes that the impact of not adopting the international 
standard would be significantly greater. Therefore, the NRC is 
proposing to adopt the radionuclide exemption values to assure 
continued consistency between domestic and international regulations.
    In Sec. 71.10(b)(3), the 0.74-TBq (20-Ci) exemption for special 
form americium and special form plutonium would be removed, except for 
\244\ Pu. This provision was originally provided in Part 71 to permit 
the transportation, in domestic commerce within the United States, of 
well-logging sealed sources containing up to 0.74 TBq (20 Ci) of 
radioactive material in Type A packages, even though that quantity of 
special form americium or plutonium was greater than the individual 
A1 limits for these radionuclides. However, over time, the 
A1 limits have been raised so that currently only \244\ Pu 
has an A1 limit less than 0.74 TBq (20 Ci) (i.e., 0.4 TBq or 
10.8 Ci). Consequently, this exemption is unnecessary for special form 
americium and special form plutonium, but is still needed for \244\ Pu.
    To prevent an unnecessary economic impact on industry, NRC staff 
believes the 0.74-TBq (20-Ci) exemption for special form \244\ Pu, 
transported in domestic commerce, should be retained as a new 
Sec. 71.14(b)(2). Furthermore, an exception would be added to 
Sec. 71.14(b)(1) indicating that paragraph (b)(1) does not apply to 
special form \244\ Pu transported in domestic commerce. This exception 
to the exemption would provide regulatory consistency between 
paragraphs (b)(1) and (b)(2), while permitting the continued 
transportation, within the U.S. only, of well-logging sources in a Type 
A package--when the source contains more than an A1 quantity 
of \244\ Pu, but less than 0.74 TBq (20 Ci). For international 
shipments, the A1 quantity limit for special form \244\ Pu 
would continue to apply.
    The NRC would include the TS-R-1 exemption values in a new table in 
Appendix A (Table A-2). Additionally, NRC recognized that changes were 
also required to Appendix A. Specifically, changes would be needed to 
paragraph II to correct the following problems: (1)

[[Page 21398]]

The existing paragraph is not in plain language; (2) Guidance is needed 
on how to determine exempt material activity concentrations and exempt 
consignment activity limits for unlisted radionuclides; (3) The method 
of requesting Commission approval, if new Table A-3 is not used, needs 
to be specified; and (4) The existing requirement on requesting NRC 
prior approval is not listed in the approved Information collection 
requirements of Sec. 71.6.
    The NRC draft RA indicates that adopting the radionuclide-specific 
exemption values contained in TS-R-1 is appropriate from a safety, 
regulatory, and cost perspective. Adoption of these values would 
provide a consistent level of protection for all radionuclides and 
result in enhanced regulatory efficiency for the NRC and consistency 
among NRC, IAEA, and DOT. In addition, adoption would result in a 
single system for determining if materials are subject to domestic or 
international regulations (e.g., an imported package from England or 
France, which is exempt, would also be exempt in the United States). 
NRC believes that this increase in regulatory efficiency and potential 
cost savings, in some cases, more than offsets the potential increased 
costs to industry. These costs are anticipated to include minor 
administrative and procedural changes to use radionuclide-specific 
exemptions. Also, industry would expend resources to identify the 
radionuclides in a material, measure the activity concentration of each 
radionuclide, and apply the ``mixture rule'' to ensure that a material 
is exempt. This is in contrast to the current approach of verifying 
that the material's total concentration is less than 70 Bq/g (0.002 
Ci/g). Further, because some low-level materials may be newly 
brought into the scope of the regulations, some additional costs may be 
incurred. However, NRC believes that these costs would be offset by the 
fact that some materials may be moved outside the scope of the 
regulations, resulting in a cost savings. Cost savings for shippers of 
low-level materials shipping both domestically and internationally 
would also be decreased because they would only have to ensure 
compliance with one set of requirements as opposed to two distinctly 
separate sets of requirements. Also, nonadoption of the TS-R-1 values 
could result in significant negative cost impacts on international 
commerce. Finally, NRC does not believe that adopting these values 
would have a significant effect on the total number of shipments 
domestically or internationally. The changes would also not 
significantly affect the way these materials are handled.
    The NRC considered the above public comments and the draft RA of 
this change, and concluded that adopting the new IAEA, dose-based, 
exemption values would improve public health and safety by establishing 
a consistent dose-model application for minimizing potential dose to 
transport workers. Within the United States, DOT has the responsibility 
for regulating the classification of radioactive materials. DOT is also 
adopting the TS-R-1 exemption concentration activity and exempt 
consignment values, and the NRC is proposing to make conforming changes 
to Part 71. While these activity concentration values will impact 
certain sectors, the impact of not adopting the international standard 
would be significantly greater. By adopting the provision to allow 
natural material and ores containing NORM, which are not intended to be 
processed for the radionuclides, to have an activity 10 times the 
exemption value, the NRC believes that Part 71's impact on the mineral 
and petroleum industries will be minimized.
    NRC Proposed Position. The NRC is proposing to adopt the 
radionuclide exemption values in TS-R-1 to assure continued consistency 
between domestic and international regulations for the basic definition 
of radioactive material. This adoption into NRC regulations would not 
impact the Memorandum of Understanding (MOU) (July 2, 1979; 44 FR 
38690) between DOT and NRC. The exemptions in existing Sec. 71.10 would 
be revised to reflect the exempt concentration and exempt consignment 
values of Appendix A, Table A-2. In addition, provisions for 10 times 
applicable values would be included for NORM and other natural 
materials. These changes would conform this rule to DOT's proposed 
regulations.
    Affected Sections. Secs. 71.10, 71.88, Appendix A.
Issue 3. Revision of A1 and A2
    Background. The international and domestic transportation 
regulations use established activity values to specify the amount of 
radioactive material that is permitted to be transported in a 
particular packaging and for other purposes. These values, known as the 
A1 and A2 values, indicate the maximum activity 
that is permitted to be transported in a Type A package. The 
A1 values apply to special form radioactive material, and 
the A2 values apply to normal form radioactive material. See 
Sec. 71.4 for definitions.
    In the case of a Type A package, the A1 and 
A2 values as stated in the regulations apply as package 
content limits. Additionally, fractions of these values can be used 
(e.g., 1x10-\3\ A2 for a limited quantity of 
solid radioactive material in normal form), or multiples of these 
values (e.g., 3,000 A2 to establish a highway route 
controlled quantity threshold value).
    Based on the results from an updated Q-system (see TS-G-1.1, 
Appendix I), the IAEA has adopted new A1 and A2 
values for radionuclides listed in TS-R-1 (see paragraph 201 and Table 
I). IAEA adopted these new values based on calculations which were 
performed using the latest dosimetric models recommended by the 
International Commission on Radiological Protection (ICRP) in 
Publication 60, ``1990 Recommendations of the ICRP.'' A thorough review 
of the Q-system also included incorporation of data from updated 
metabolic uptake studies. In addition, several refinements were 
introduced in the calculation of contributions to the effective dose 
from each of the pathways considered. The pathways themselves are the 
same ones considered in the 1985 version of the Q-system: external 
photon dose, external beta dose, inhalation dose, skin and ingestion 
dose from contamination, and dose from submersion in gaseous 
radionuclides. A thorough, up-to-date radiological assessment has been 
performed for each radionuclide of potential exposures to an individual 
should a Type A package of radioactive material be involved in an 
accident during transport. The new A1 and A2 
values reflect that assessment.
    While the dosimetric models and dose pathways within the Q-system 
were thoroughly reviewed and updated, the reference doses were 
unchanged. The reference doses are the dose values which are used to 
define a ``not unacceptable'' dose in the event of an accident. 
Consequently, while some revised A1 and A2 values 
are higher and some are lower, the potential dose following an accident 
is the same as with the previous A1 and A2 
values. The revised dosimetric models are used internationally to 
calculate doses from individual radionuclides, and these refinements in 
the pathway calculations result in various changes to the A1 
and A2 values. In other words, where an A1 or 
A2 value has increased, the potential dose is still the 
same--the use of the revised dosimetric models just shows that a higher 
activity of that radionuclide is actually required to produce the same 
reference dose. Conversely, where an A1 or A2 
value has decreased, the revised models show that

[[Page 21399]]

less activity of that nuclide is needed to produce the reference dose.
    Discussion. Comments on the adoption of the new A1 and 
A2 values were received during the three public meetings and 
on the NRC website. One commenter stated that to conduct business 
internationally, there needs to be consistency between the 
international and domestic regulations. These commenters supported the 
adoption of the new values into Part 71. Other industry 
representatives, however, indicated the values should not change as 
they would need to modify the computer codes at their facility to 
maintain the ability to accurately meet the regulatory requirements for 
transportation. Other commenters were concerned about the safety 
aspects of transportation and the emergency responder's exposure if the 
new values should be adopted.
    Additional comments were received concerning the A1 and 
A2 values for californium-252 and molybdenum-99, 
respectively. Currently, in Part 71, the A1 for californium-
252 is 0.1TBq (2.7 Ci). The A1 value in TS-R-1 is 
5.0x10-\2\ TBq (1.35 Ci). Both NRC and DOT have learned that 
IAEA is considering changing the A1 value for californium-
252 back to the value currently in 10 CFR Part 71 and 49 CFR in the 
next edition of TS-R-1. DOT is proposing to retain the current Part 71 
A1 value for californium-252 for domestic commerce. 
Therefore the NRC is planning to do the same as a conforming action 
with DOT.
    Regarding molybdenum-99, comments were received from the 
radiopharmaceutical industry concerning the A2 value. 
Currently in Part 71, the A2 value for molybdenum-99 is 0.5 
TBq (13.5 Ci). Further, in Appendix A, Table A-1, the A2 
value for molybdenum-99 has a footnote that indicates for domestic use, 
the A2 value is 0.74 TBq (20 Ci). Pharmaceutical industry 
representatives indicated that a change to the TS-R-1 A2 
value of 0.6 TBq (16.2 Ci) for molybdenum-99 would result in a 
significant increase in the number of packages shipped and in 
occupational doses due to the lower A2 value (16.2 Ci versus 
20 Ci). DOT is proposing to retain the current exception for 
molybdenum-99 for domestic commerce, and NRC also believes the current 
exception for this radionuclide should be retained.
    Several commenters opposed NRC's proposal to adopt the IAEA 
A1 and A2 values, arguing that any increase in 
allowable activity levels is unacceptable, could result in increased 
risk, and would violate the principle of maintaining safety. One 
commenter stated that the proposed adoption would change from an 
activity-based limit system to a dose-based limit system, which is 
unacceptable because dose-based limits are more difficult to verify and 
enforce than are activity-based limits.
    Several commenters stated that NRC should provide a breakdown of 
which radionuclides would have increased activity levels, and which 
would remain the same, to allow for meaningful public comment on the 
proposed change.
    Several commenters indicated that adoption of ICRP-60 into NRC 
regulations would result in another inconsistency within the 
regulations. Another commenter disagreed, arguing that NRC runs the 
risk of eroding public confidence in its regulatory role by accepting, 
then ignoring, the advice of international experts. The commenter 
argued that there should be a very strong justification if 
recommendations of the ICRP are to be discounted.
    In general, the new A1 and A2 values are 
within a factor of about three of the earlier values; there are a few 
radionuclides where the new A1 and A2 values are 
outside this range. A few tens of radionuclides (out of more than 300) 
have new A1 values higher than previous values by factors 
ranging between 10 and 100. This is due mainly to improved modeling for 
beta emitters. There are no new A1 or A2 values 
that are lower than the previous figures by more than a factor of 10. A 
few radionuclides previously listed are now excluded, but two 
additional ones have been added, both isomers of europium-150 and 
neptunium-236. Many A1 and A2 values remain 
unchanged.
    The NRC staff review of TS-R-1 against the current Part 71 has 
identified 16 radionuclides that are listed in Table A-1 in Part 71 
Appendix A, but which do not appear in TS-R-1. These are: Ar-42, Au-
196, Es-253, Es-254, Es-254m, Es-255, Fm-255, Fm-257, Ho-163, Ir-193m, 
Nb-92m, Po-208, Po-209, Re-183, Te-118, and Tm-168. In an effort to 
maintain compatibility with TS-R-1, the NRC proposes not to include 
A1 and A2 values for these radionuclides in Table 
A-1. Licensees can use, without NRC approval, the general values for 
A1 or A2 in Table A-3 for individual 
radionuclides whose identities are known (such as the above 16), but 
which are not listed in Table A-1. Alternatively, licensees can obtain 
NRC approval for using specific values for those radionuclides. The NRC 
staff consulted with the DOT staff on this issue, and DOT is also 
proposing not to include A1 and A2 values for 
these radionuclides in its revised table of A1 and 
A2 values.
    The A1 and A2 values were revised by IAEA 
based on refined modeling of possible doses from radionuclides. The NRC 
staff believes adoption of the IAEA standard would be an overall 
benefit to public and worker health and international commerce by 
ensuring that the A1 and A2 values are consistent 
within and between international and domestic transportation 
regulations.
    The NRC draft RA indicates that adopting the new A1 and 
A2 activity limits specified in TS-R-1 is appropriate from a 
safety, regulatory, and cost perspective. Adoption of these values 
would result in enhanced regulatory efficiency for the NRC and 
consistency among NRC, IAEA, and DOT, especially in the handling of 
imports and exports. Adoption would result in a single set of values 
for determining the activity limits for specifying the amount of 
radioactive material permitted to be transported in a particular 
package for both domestic and international shipments. In some cases, 
NRC believes that this increase in regulatory efficiency and potential 
cost savings more than offsets the potential increased costs. These 
costs are anticipated to include revisions to shipping programs to 
implement the new values, modifications to shipping processes to assure 
compliance with the new values, and training. These costs, however, are 
expected to be minor because industry already has programs in place 
that use the A1 and A2 values. In addition, NRC 
would realize additional minor implementation costs in revising the 
values in Part 71. The NRC draft RA indicated no significant change in 
the number of shipments per year; therefore, accident frequency would 
not be affected.
    NRC Proposed Position. The NRC is proposing to make a conforming 
change to Part 71 to adopt the new A1 and A2 
values from TS-R-1 in Part 71, with the differences as discussed for 
molybdenum-99 and californium-252. The NRC is also proposing not to 
include A1 and A2 values for the 16 radionuclides 
that are currently listed in Part 71, but which do not appear in TS-R-1 
(see the Discussion section of Issue 3). This action would allow for 
continued consistency within and between international and domestic 
transportation regulations for radioactive materials. The DOT is also 
proposing to adopt the new TS-R-1 A1 and A2 
values in its regulations, but without the 16 radionuclides cited 
above. NRC is requesting stakeholder input with regard to the changes 
focused around the A1 and A2 values for 
californium-252, molybdenum-99, and the 16 radionuclides that will be

[[Page 21400]]

removed from Table A-1. NRC is interested in learning what impacts 
these changes will have on industry.
    Affected Sections. Appendix A.
Issue 4. Uranium Hexafluoride Package Requirements
    Background. Requirements for uranium hexafluoride (UF6) 
packaging and transportation are found in both NRC and DOT regulations. 
The DOT regulations contain requirements that govern many aspects of 
UF6 packaging and shipment preparation, including a 
requirement that the UF6 material be packaged in cylinders 
that meet the ANSI N14.1 standard. NRC regulations address fissile 
materials and Type B packaging designs for all materials.
    TS-R-1 contains detailed requirements for UF6 packages 
designed for transport of more more than 0.1 kg UF6. First, 
TS-R-1 requires the use of the International Organization for 
Standardization (ISO) 7195, ``Packaging of Uranium Hexafluoride for 
Transport.'' Second, TS-R-1 requires that all packages containing more 
than 0.1 kg UF6 must meet the ``normal conditions of 
transport'' drop test, a minimum internal pressure test, and the 
hypothetical accident condition thermal test (para 630). However, TS-R-
1 does allow a competent national authority to waive certain design 
requirements, including the thermal test for packages designed to 
contain greater than 9,000 kg UF6, provided that 
multilateral approval is obtained. Third, TS-R-1 prohibits 
UF6 packages from using pressure relief devices (para 631). 
Fourth, TS-R-1 includes a new exception for UF6 packages 
regarding the evaluation of criticality safety of a single package. 
This new exception (para 677(b)) allows UF6 packages to be 
evaluated for criticality safety without considering the inleakage of 
water into the containment system. Consequently, a single fissile 
UF6 package does not have to be subcritical assuming that 
water leaks into the containment system. This provision only applies 
when there is no contact between the valve body and the cylinder body 
under accident tests, and the valve remains leak-tight, and when there 
are quality controls in the manufacture, maintenance, and repair of 
packagings coupled with tests to demonstrate closure of each package 
before each shipment.
    Discussion. One commenter indicated serious concerns about the 
safety margins for UF6 packaging. The commenter cited the 
exception in TS-R-1, paragraph 677(b), which would allow UF6 
packages to be evaluated for criticality without considering the 
inleakage of water. The commenter cited a report describing one case 
where UF6 packages with manufacturing defects were used. The commenter 
indicated that it would be imprudent and unwise public policy to assume 
that water could not leak into a package containing UF6.
    Another commenter stated that a justification for the reduced 
regulatory burden has not been established and cannot be done unless a 
risk study, which determines the level of conservatism currently 
contained in Part 71, is conducted. Without this analysis, the 
commenter argued, reduction of regulatory burden leading to inadvertent 
criticality could lead to loss of life, degradation of the environment, 
economic repercussions, and degradation of public confidence.
    Also, comments at the public meetings supported the NRC view that 
ANSI N14.1 and ISO 7195 are equivalent. Further, other comments 
indicated that NRC-certified UF6 packages already comply 
with TS-R-1 paragraphs 630 and 677(b).
    The provisions of Sec. 71.55(b) specify that a fissile material 
package must be designed, or the contents limited, so that a single 
package would be critically safe if water were to leak into the 
containment vessel. This is a design feature that assures criticality 
safety in transport, in the unanticipated event that water leaks into 
the containment vessel, and provides moderating materials for the 
fissile contents. The proposed new Sec. 71.55(g) would except fissile 
UF6 from the requirement that a single package must be 
critically safe with water inleakage. This is consistent with the 
worldwide practice in shipping fissile UF6 and is consistent 
with ANSI N14.1 and ISO 7195 standards and DOT regulations.
    The proposed rule language further restricts use of the exception 
to a maximum enrichment of 5 weight percent uranium-235. This is the 
maximum enrichment currently authorized in ANSI N14.1, ISO 7195, and 
DOT regulations in cylinders larger than 20.3 cm (8 inches) in 
diameter. For smaller cylinders, the exception is not needed because 
current enrichments are critically safe by geometry for a single 
package. The exception, with the enrichment limit, codifies current 
worldwide practice in shipping fissile uranium hexafluoride. Large 
quantities of enriched (greater than 5 weight percent uranium-235) 
UF6 would require packages that meet the water inleakage 
standards in Sec. 71.55(b). The staff believes that it is not prudent 
to expand this exception to include UF6 shipments with 
higher uranium enrichments.
    The NRC draft RA indicates that revising the current requirements 
for uranium hexafluoride packages to include an exception from the 
requirement that single packages must be critically safe from water 
inleakage is appropriate from a safety, regulatory, and cost 
perspective. In developing the draft RA, the NRC first determined that 
there are no substantial differences between ANSI N14.1 standard and 
ISO 7195 standard for UF6 packaging, and therefore, there 
would be no significant cost impacts from this change, because NRC 
currently requires conformance with ANSI N14.1, but regulatory 
efficiency would be enhanced by making Part 71 compatible with TS-R-1. 
The internal pressure test and drop test requirements are currently met 
by existing package designs that comply with ANSI N14.1. Therefore, 
there would be limited impact on licensees by this aspect of the NRC 
action. The NRC staff also considered the United States' earlier 
opposition (Taylor, 1996) to this change, i.e., the IAEA adopting the 
UF6 package requirements. Most of the impact of adopting the 
TS-R-1 UF6 provisions would fall on the 30-inch and 48-inch 
bare cylinders that are within the purview of DOT and for which there 
is a ``multilateral'' approval option that could be used to mitigate 
most of this potential impact to licensees. Therefore, the adoption of 
the TS-R-1 requirements is not expected to have significant impact on 
fissile package designs for UF6. (Additional minor costs may 
be incurred for training for handling overpacks.) Because the changes 
are not expected to have significant impacts on current package 
designs, changes in environmental impacts are expected to be 
negligible.
    NRC Proposed Position. The NRC is proposing to adopt Sec. 71.55(g) 
to address TS-R-1, paragraph 677(b), to exempt certain UF6 
packages from the requirements of Sec. 71.55(b). The requirements in 
TS-R-1, paragraphs 629, 630, and 631, do not necessitate changes to 
Part 71 because NRC uses analogous national standards and addresses 
package design requirements in its design review process. All NRC-
certified packages must be used in accordance with DOT requirements 
(including the UF6 requirement in 49 CFR 173.420).
    Affected Sections. Sec. 71.55.
Issue 5. Introduction of the Criticality Safety Index Requirements
    Background. Historically, the IAEA and U.S. regulations (both NRC 
and DOT) have used a term known as the

[[Page 21401]]

Transport Index (TI) to determine appropriate safety requirements 
during transport. TI has been used to control the accumulation of 
packages for both radiological safety and criticality safety purposes 
and to specify minimum separation distances from persons (radiological 
safety). The TI has been a single number which is the larger of two 
values: the ``TI for criticality control purposes''; and the ``TI for 
radiation control purposes.'' Taking the larger of the two values has 
ensured conservatism in limiting the accumulation of packages in 
conveyances and in-transit storage areas.
    TS-R-1 (paragraph 218) has introduced the concept of a Criticality 
Safety Index (CSI) separate from the old TI. As a result, the TI was 
redefined in TS-R-1. The CSI is determined in the same way as the ``TI 
for criticality control purposes,'' but now it must be displayed on 
shipments of fissile material (paragraphs 544 and 545) using a new 
``fissile material'' label. The redefined TI is determined in the same 
way as the ``TI for radiation control purposes'' and continues to be 
displayed on the traditional ``radioactive material'' label.
    TS-R-1 (paragraph 530) also increased the allowable per package TI 
limit [for criticality control purposes (new CSI)] from 10 to 50 for 
nonexclusive use shipments. No change was made to the per package 
radiation TI limit of 10 for nonexclusive use shipments. As noted 
above, a consolidated radiation safety and criticality safety index 
existed in the past. In this consolidated index, the per package TI 
limit of 10 was historically based on concerns regarding the fogging of 
photographic film in transit, because film might also be present on a 
nonexclusive use conveyance. Consequently, when the single radiation 
and criticality safety indexes were split into the TI and CSI indexes, 
the IAEA determined that the CSI per package limit, for fissile 
material packages that are shipped on a nonexclusive use conveyance, 
could be raised from 10 to 50. The IAEA believed that limiting the 
total CSI to less than or equal to 50 in a nonexclusive use shipment 
provided sufficient safety margin, whether the shipment contains a 
single package or multiple packages. Therefore, the per package CSI 
limit, for nonexclusive use shipments, can be safely raised from 10 to 
50, thereby providing additional flexibility to shippers. Additionally, 
no change was made to the per package CSI limit of 100 for exclusive 
use shipments.
    Discussion. Comments received on this proposal indicated that the 
industry supports the use of the new label ``CSI'' in conjunction with 
the ``TI'' labels, and stated that separate labels are more meaningful 
and provide additional safety in transport, as long as the two labels 
are distinctive, so as to avoid confusion.
    In general, public comments received at the meetings supported the 
use of the CSI. One commenter believed that using the TI as the means 
to control criticality safety does not provide emergency responders 
with information on the undamaged condition of the package. Other 
commenters suggested that NRC should provide the underlying technical 
justification for the term ``equivalent safety,'' because otherwise, 
this change would seemingly allow for more packages in a single 
shipment. The use of CSI provides an equivalent level of safety to 
using a TI, because the CSI uses the same methodology (Sec. 71.59) that 
was used to calculate the criticality portion of the current TI.
    One industry commenter disagreed that the CSI requirement is 
appropriate. The commenter stated that the TI already incorporates the 
more restrictive value and provides adequate protection. The commenter 
believed there is no increase in safety by adding this new requirement 
and, in fact, it would result in more opportunities for human error. 
Further, the commenter indicated that any benefit for adding the CSI is 
far outweighed by the additional labor, material, training, and 
administration costs that would be borne by a company that ships 
thousands of packages each year.
    Increasing the CSI per package limit from 10 to 50 for nonexclusive 
use shipments was overlooked by NRC staff and was not discussed in the 
June 2000 Issues Paper or the associated public meetings. Consequently, 
no stakeholder input was obtained on this aspect of Issue 5 prior to 
developing the proposed rule.
    The NRC draft RA indicates that introducing new CSI requirements 
into part 71 is appropriate from a safety, regulatory, and cost 
perspective. NRC would require that applicants for fissile material 
package design approvals clearly indicate the CSI value for the design. 
The CoCs the NRC issues for these designs would also need to clearly 
indicate the CSI value for authorized contents. The adoption of the CSI 
values would make part 71 consistent with TS-R-1, therefore enhancing 
regulatory efficiency.
    The NRC staff believes that shipping fissile material packages on 
either an exclusive or nonexclusive use conveyance provides a 
reasonable assurance that public health and safety and the environment 
will be adequately protected. Furthermore, shipment on a nonexclusive 
use conveyance of a single package with a CSI equal to 50, a shipment 
of 5 packages each with a CSI equal to 10, or 20 packages each with a 
CSI equal to 2.5, are all safe and provide reasonable assurance of 
adequate protection. While NRC staff recognizes that the reactivity per 
package will increase with an increase in the CSI from 10 to 50, staff 
also believes the limit on the total CSI in a nonexclusive use shipment 
provides adequate protection against mishandling events. Accordingly, 
this change will not have a significant safety impact.
    The total annual estimated cost of the new label to the nuclear 
power licensees and material licensees is approximately $1.4 
million.\1\ Some of these costs would be offset by the fact that for 
some shipments of fissile material packages, the accumulation of 
packages for criticality control purposes and the accumulation of 
packages (including minimum separation distances from persons) for 
radiological control purposes are shipped independently (the most 
restrictive criteria would not control the other as is the case with 
the current dual-use TI). Further, increased efficiency in shipping 
some fissile material packages could occur by avoiding the situation 
where separation distance requirements (radiological safety) unduly 
restrict package accumulation (criticality safety). From a health and 
safety perspective, emergency responders in accident circumstances 
(thus public health and safety) benefit from more clearly displayed 
information upon arrival at the accident scene.
---------------------------------------------------------------------------

    \1\ This number is estimated by assuming 10 percent of the 
approximately 2.8 million total annual shipments (or 280,000) 
contain fissile material requiring lables indicating the CSI and TI. 
And of this 10 percent, NRC assumes five packages per shipment and 
$1 per package for labeling, thus arriving at the $1.4 million total 
annual licensee costs.
---------------------------------------------------------------------------

    The NRC staff was unable to estimate the magnitude of the impact or 
cost savings that would arise to licensees due to the increase in the 
CSI per package limit. However, staff judged that cost savings could be 
realized because of increased licensee flexibility in shipping a larger 
number of fissile material packages on less expensive, nonexclusive use 
conveyances. Therefore, the NRC is requesting stakeholder input on the 
quantity of shipments in a typical year that would be affected by an 
increase in the per package CSI limit from 10 to 50 for nonexclusive 
use shipments and any associated cost savings. Because of lack

[[Page 21402]]

of data, the NRC is also requesting stakeholder input on the current 
number of fissile material shipments typically made per year (i.e., 
fissile-exempt, fissile general license, or Type A(F) or B(F) 
packages); the types of material shipped (e.g., waste, laboratory 
quantities, or production quantities); the shipment method used for 
these types of fissile material; and whether these are exclusive or 
nonexclusive use shipments.
    NRC Proposed Position. The NRC proposes to adopt the TS-R-1 
(paragraph 218) which incorporates a CSI in Part 71 that would be 
determined in the same manner as the current Part 71 ``TI for 
criticality control purposes.'' The NRC also proposes to adopt TS-R-1 
(paragraph 530) which increases the CSI per package limit from 10 to 50 
for fissile material packages in nonexclusive use shipments. A TI will 
be determined in the same way as the ``TI for radiation control 
purposes.'' The NRC believes the differentiation between criticality 
control and radiation protection would better define the hazards 
associated with a given package and, therefore, provide better package 
hazard information to emergency responders. The increase in the per 
package CSI limit may provide additional flexibility to licensees by 
permitting the increased use of less-expensive, nonexclusive use 
shipments. However, licensees will still retain the flexibility to ship 
a larger number of packages of fissile material on an exclusive use 
conveyance.
    Affected Sections. Secs. 71.4, 71.18, 71.20, 71.59.
Issue 6. Type C Packages and Low Dispersible Material
    Background. TS-R-1 has introduced two new concepts: the Type C 
package (paragraphs 230, 667-670, 730, 734-737) and the Low Dispersible 
Material (LDM). The Type C packages are designed to withstand severe 
accident conditions in air transport without loss of containment or 
significant increase in external radiation levels. The LDM has limited 
radiation hazard and low dispersibility; as such, it could continue to 
be transported by aircraft in Type B packages (i.e., LDM is excepted 
from the TS-R-1 Type C package requirements). U.S. regulations do not 
contain a Type C package or LDM category, but do have specific 
requirements for the air transport of plutonium (Secs. 71.64 and 
71.74). These specific NRC requirements for air transport of plutonium 
would continue to apply.
    The Type C requirements apply to all radionuclides packaged for air 
transport that contain a total activity value above 3,000 A1 
or 100,000 A2, whichever is lesser, for special form 
material, or above 3,000 A2 for all other radioactive 
material . Below these thresholds, Type B packages would be permitted 
to be used in air transport. The Type C package performance 
requirements are significantly more stringent than those for Type B 
packages. For example, a 90-meter per second (m/s) impact test is 
required instead of the 9-meter drop test. A 60-minute fire test is 
required instead of the 30-minute requirement for Type B packages. 
There are other additional tests, such as a puncture/tearing test, 
imposed for Type C packages. These stringent tests are expected to 
result in package designs that would survive more severe aircraft 
accidents than Type B package designs.
    The LDM specification was added in TS-R-1 to account for 
radioactive materials (package contents) that have inherently limited 
dispersibility, solubility, and external radiation levels. The test 
requirements for LDM to demonstrate limited dispersibility and 
leachability are a subset of the Type C package requirements (90-m/s 
impact and 60-minute thermal test) with an added solubility test, and 
must be performed on the material without packaging for nonplutonium 
materials. The LDM must also have an external radiation level below 10 
mSv/hr (1 rem/hr) at 3 meters. Specific acceptance criteria are 
established for evaluating the performance of the material during and 
after the tests (less than 100 A2 in gaseous or particulate 
form of less than 100-micrometer aerodynamic equivalent diameter and 
less than 100 A2 in solution). These stringent performance 
and acceptance requirements are intended to ensure that these materials 
can continue to be transported safely in Type B packages aboard 
aircraft.
    In 1996, the NRC communicated to the IAEA that the NRC did not 
oppose the IAEA adoption of the newly created Type C packaging 
standards (letter dated May 31, 1996, from James M. Taylor, EDO, NRC, 
to A. Bishop, President, Atomic Energy Control Board, Ottawa, Canada). 
However, Mr. Taylor stated in the letter that to be consistent with 
U.S. law, any plutonium air transport to, within, or over the U.S. will 
be subject to the more rigorous U.S. packaging standards.
    Discussion: Comments from the public suggested that Type C 
standards might increase the number of shipments with smaller 
quantities of material using the same Type B containers to avoid the 
cost of developing Type C packages and to avoid the requirement of 
meeting the new Type C package standards. One commenter indicated that 
any proposal to change package design requirements should only be 
contemplated after a thorough technical review that has independently 
justified the change as protective.
    However, one commenter stated that NRC should remove from its 
regulations the plutonium-specific requirements for air transport, and 
replace them with the Type C package requirements. Also, the commenter 
stated that because Type C package development would take a number of 
years, industry would work with the NRC to define tests, analyses, and 
criteria for demonstrating compliance with the Type C package 
standards.
    One commenter questioned the rigorousness of the testing described 
in TS-R-1, indicating that the minimum acceptable impact speed should 
be increased to at least 129 m/s, as was mandated by Congress.
    The staff evaluated the Type C package, and proposes that the NRC 
not adopt Type C or LDM requirements at this time. The bases for this 
staff proposal include: (1) IAEA development of aircraft accident 
severity information through a coordinated research project for further 
evaluation of the Type C and LDM requirements; (2) the fact that there 
are very few anticipated shipments affected by these requirements; (3) 
DOT rules that permit the use of IAEA standards in nonplutonium import/
export shipments of foreign certified Type C containers, so that 
international commerce is not impacted; (4) NRC's domestic regulations 
currently in place (Secs. 71.64 and 71.74), based on specific statutory 
mandates, governing air transport of plutonium (plutonium air transport 
was a considerable factor in IAEA adoption of Type C provisions); and 
(5) comments made by the public on the issues which generally disagreed 
with or questioned the rigor of the Type C tests, and supported NRC 
maintaining its current regulatory requirements for the safety of 
plutonium air shipments.
    The DOT reviews the use of packages for import or export shipment. 
Consequently, foreign Type C packages could be approved by DOT for 
import and export only. The NRC does not believe that a Type C package 
is needed for domestic commerce; therefore, no provisions would be 
added to Part 71 relating to Type C packages. However, should DOT 
request that NRC perform a technical evaluation for a revalidation of a 
foreign Type C package design, NRC would evaluate the design against 
TS-R-1 Type C standards. Similarly, if requested by DOT, NRC would 
review a domestic Type C package design intended for use in 
international

[[Page 21403]]

commerce against TS-R-1, and provide NRC's recommendation to DOT. (Note 
that NRC revalidation of designs for DOT does not constitute NRC 
issuance of a CoC.)
    The NRC draft RA indicates that not adopting the TS-R-1 Type C or 
LDM provisions in Part 71 is appropriate from a safety, regulatory, and 
cost standpoint. There may be some reduction in regulatory efficiency 
as a result of the nonadoption of the TS-R-1 requirements, which could 
result in NRC case-by-case reviews to support international shipments. 
NRC would continue to use its proven, safe regulatory requirements for 
air transport of plutonium. Further, NRC staff resources are conserved 
by nonadoption, and no additional costs would be incurred by industry. 
Any additional costs to industry would involve development costs for 
the design of new packages to meet the Type C requirements rather than 
using existing Type B packages.
    NRC Proposed Position. The NRC would not adopt Type C or LDM 
requirements at this time.
    Affected Sections. None (not adopted).
Issue 7. Deep Immersion Test
    Background. TS-R-1 expanded the performance requirement for the 
deep water immersion test (paragraphs 657 and 730) from the 
requirements in the IAEA Safety Series No. 6, 1985 edition. Previously, 
the deep immersion test was only required for packages of irradiated 
fuel exceeding 37 PBq (1,000,000 Ci). The deep immersion test 
requirement is found in Safety Series No. 6, paragraphs 550 and 630, 
and basically stated that the test specimen be immersed under a head of 
water of at least 200 meters (660 ft) for a period of not less than one 
hour, and that an external gauge pressure of at least 2 MPa (290 psi) 
shall be considered to meet these conditions. The TS-R-1 expanded 
immersion test requirement (now called enhanced immersion test) now 
applies to all Type B(U) [Unilateral] and B(M) [Multilateral] packages 
containing more than 10\5\ A2, as well as Type C packages.
    In its September 28, 1995 (60 FR 50248), rulemaking for Part 71 
compatibility with the 1985 edition of Safety Series No. 6, the NRC 
addressed the new Safety Series No. 6 requirement for spent fuel 
packages by adding Sec. 71.61, ``Special requirements for irradiated 
nuclear fuel shipments.'' Currently, Sec. 71.61 is more conservative 
than Safety Series No. 6 with respect to irradiated fuel package design 
requirements. It requires that a package for irradiated nuclear fuel 
with activity greater than 37 PBq (10\6\ Ci) must be designed so that 
its undamaged containment system can withstand an external water 
pressure of 2 MPa (290 psi) for a period of not less than one hour 
without collapse, buckling, or inleakage of water. The conservatism 
lies in the test criteria of no collapse, buckling, or inleakage as 
compared to the ``no rupture'' criteria found in Safety Series No. 6 
and TS-R-1. The draft advisory document for TS-R-1 (TS-G-1.1, 
paragraphs 657.1 to 657.7) recognizes that leakage into the package and 
subsequent leakage from the package are possible while still meeting 
the IAEA requirement.
    The Safety Series No. 6 test requirements were based on risk 
assessment studies that considered the possibility of a ship carrying 
packages of radioactive material sinking at various locations. The 
studies found that, in most cases, there would be negligible harm to 
the environment if a package were not recovered. However, should a 
large irradiated fuel package (or packages) be lost on the continental 
shelf, the studies indicated there could be some long term exposure to 
man through the food chain. The 200-meter (660-ft) depth specified in 
Safety Series No. 6 is equivalent to a pressure of 2 MPa (290 psi), and 
roughly corresponds to the continental shelf and to depths that the 
studies indicated radiological impacts could be important. Also, 200 
meters (660 ft) was a depth at which recovery of a package would be 
possible, and salvage would be facilitated if the containment system 
did not rupture. (Reference Safety Series No. 7, paragraphs E-550.1 
through E-550.3.)
    The expansion in scope of the deep immersion test was due to the 
fact that radioactive materials, such as plutonium and high-level 
radioactive wastes, are increasingly being transported by sea in large 
quantities. The threshold defining a large quantity as a multiple of 
A2 is considered to be a more appropriate criterion to cover 
all radioactive materials, and is based on a consideration of potential 
radiation exposure resulting from an accident.
    Discussion. Several comments received at the public meetings, as 
well as written comments received on the Issues Paper, indicated 
support for retaining the current, more stringent, requirements 
contained in Sec. 71.61 with respect to not allowing collapse, 
buckling, or inleakage of water in the containment vessel. One 
commenter was concerned that the term ``rupture'' seemed less stringent 
than ``collapse, buckling, or inleakage of water.'' The commenter 
noted, however, that the issues paper does not include definitions for 
``rupture'' or ``buckling,'' so it is difficult to know which term is 
more or less stringent. Another commenter believed that the proposed 
test requirement of withstanding underwater pressure for at least an 
hour is insufficient. The commenter explained that it is unrealistic to 
expect to recover nuclear materials from the water within 1 hour after 
a major accident.
    One commenter questioned whether there was sufficient technical 
justification for relaxing the current NRC test criteria for packages 
of irradiated nuclear fuel. The commenter stated that a lot of 
environmental damage can occur before a rupture develops, and that the 
proposal does nothing to ensure that packages are as safe as they can 
be.
    Another commenter noted that TS-R-1 refers only to normal form 
material for the immersion test. Specifically, the commenter asked what 
the criteria are for a special form A1 quantity, and whether 
the deep immersion test was necessary for B(U) packages for special 
form materials. NRC reviewed the IAEA regulations and believes that 
this requirement applies to both normal form and special form material. 
Similarly, one commenter noted that, in practicality, the quantities 
listed would be limited to irradiated fuel elements, and that shipment 
of radioisotopes rarely contain these amounts. This commenter suggested 
that the present criteria be maintained and extended to cover all 
packages with activity levels greater than or equal to 105 
A2 quantities with the note that this is more conservative 
than TS-R-1 requirements. The commenter stated this should eliminate 
the requirement for special review and certification of U.S. origin 
package designs. For nonirradiated fuel element shipments, the 
commenter believed there should be no impact on availability and 
shipping costs because there are few shipments of the required 
quantities of this material. Finally, the commenter questioned whether, 
with the application to B(U) packages containing A1 special 
form sources, these packages are exempt from this test.
    In response to the question about how to address the differences in 
acceptance standards, two commenters stated that due to the 
international nature of transportation activities, U.S. transportation 
regulations should be consistent with IAEA transportation regulations 
and, therefore, NRC should adopt the TS-R-1 requirements for the 
enhanced deep immersion test.
    Two commenters also addressed whether U.S. origin package designs

[[Page 21404]]

should be specifically reviewed and certified before shippers can 
export them. One commenter said that if the response is not specific to 
the deep immersion test, but applies to all package design criteria, 
then the shipment of U.S. certified package designs for import/export 
use beginning in mid-2001 is entirely dependent upon approval of these 
designs to TS-R-1 performance standards. The commenter believed that 
failure to grant U.S. Competent Authority certifications for these 
designs would seriously hinder the industrial radiography industry, and 
place U.S. package designers and manufacturers at a strong competitive 
disadvantage. The commenter added that several of its shipments were 
not acceptable in several countries when NRC and DOT failed to adopt 
Safety Series No. 6 in a timely manner.
    Another commenter stated that NRC should clarify if previously 
approved packages would be grandfathered, or if they would have to be 
recertified by means of a deep immersion test.
    The NRC proposes revising Part 71 requiring an enhanced water 
immersion test for packages used for radioactive contents with activity 
greater than 105 A2. Section 71.61 currently 
refers to packages for irradiated fuel with activity greater than 37 
PBq (106 Ci); the water immersion test would need to be 
changed to apply to Type B packages containing greater than 
105 A2 and Type C packages. Given that any 
package containing spent fuel with activity greater than 37 PBq 
(106 Ci) would also have an activity significantly greater 
than 105 A2, such a change would bound Type B 
spent fuel packages currently addressed in 10 CFR 71.61. Therefore, a 
specific reference to special requirements for irradiated nuclear fuel 
shipments would no longer be required.
    As mentioned earlier, there is a difference between the test 
acceptance criteria specified in TS-R-1 and Sec. 71.61. Safety Series 
No. 6 refers to no rupture, while Sec. 71.61 requires no collapse, 
buckling, or inleakage of water when subjected to the test conditions. 
In the September 28, 1995, rulemaking, NRC staff provided justification 
for the more specific NRC acceptance criteria. The rulemaking stated 
that: ``NRC has since determined that the term `rupture' cannot be 
determined by engineering analysis and that NRC has decided to change 
the acceptance criteria for the deep immersion test from `rupture' to 
`collapse, buckling, or inleakage of water'.'
    Given that the TS-R-1 background material does not provide any new 
information on defining the term ``rupture'' from that provided for 
Safety Series No. 6, the NRC intends to retain the current 
interpretation of ``rupture'' to mean ``collapse, buckling, or 
inleakage of water,'' in any revision to Sec. 71.61. During the comment 
period for the proposed rule, should information be provided about how 
the term ``rupture'' should be defined, or on how foreign countries 
have certified packages to this criterion, then the NRC will consider 
this in determining whether the ``collapse, buckling, or inleakage of 
water'' criteria should be revised before issuing the final rule.
    The NRC draft RA indicates that revising Part 71 to require an 
enhanced water immersion test for packages used for radioactive 
contents with activity greater than 105 A2 while 
retaining the current Sec. 71.61 interpretation of ``rupture'' to mean 
``collapse, buckling, or inleakage of water,'' is appropriate from a 
safety, regulatory, and cost perspective. The proposed change would 
improve regulatory efficiency by bringing U.S. regulations in harmony 
with the standards contained in TS-R-1. This would improve the 
efficiency of handling imports and exports and would make U.S. 
standards compatible with other IAEA Members States.
    Implementation of the proposed change could result in costs to 
licensees as they test and certify packages to the proposed standard. 
The NRC may incur costs for developing procedures, reviewing and 
approving test results, and recertifying packages. The proposed change 
may reduce impacts to public health in the case of an accident. A 
package tested to the new requirements would be able to withstand 
pressure at increased depths without collapsing, buckling, or allowing 
inleakage of water, thereby keeping the radioactive materials enclosed. 
The likelihood of a member of the public receiving a dose from a 
package resting in deep water is exceedingly small and would be even 
smaller if the proposed change were implemented in that the test would 
apply to a broad range of packages. Moreover, the duration of the test, 
1 hour, is reasonable for a package resting in deep water, because the 
water pressure will be constant, and the 1-hour test will clearly 
establish if the package can withstand that pressure. A successfully-
tested package would be able to withstand the pressure at this depth 
without rupturing, thereby keeping the radioactive materials enclosed 
and permitting a reasonable length of time for recovery. Retaining 
package integrity would prevent the possible expenses of restricting 
the area (to prevent users such as boaters or fishers from entering the 
vicinity) and remediating any contamination of the marine environment.
    NRC Proposed Position. The NRC proposes to adopt the requirement 
for enhanced water immersion test for packages used for radioactive 
contents with activity greater than 105 A2. The 
NRC intends to retain the current test requirements in Sec. 71.61 of 
``one hour without collapse, buckling, or inleakage of water.''
    Affected Sections. Secs. 71.41, 71.51, 71.61.
Issue 8. Grandfathering Previously Approved Packages
    Background. Historically, the IAEA, DOT, and NRC regulations have 
included transitional arrangements or ``grandfathering'' provisions 
whenever the regulations have undergone major revision. The purpose of 
grandfathering is to minimize the costs and impacts of implementing 
changes in the regulations on existing package designs and packagings. 
Grandfathering typically includes provisions that allow: (1) Continued 
use of existing package designs and packagings already fabricated, 
although some additional requirements may be imposed; (2) completion of 
packagings that are in the process of being fabricated or that may be 
fabricated within a given time period after the regulatory change; and 
(3) limited modifications to package designs and packagings without the 
need to demonstrate full compliance with the revised regulations, 
provided that the modifications do not significantly affect the safety 
of the package.
    Each transition from one edition of the IAEA regulations to another 
(and the corresponding revisions of the NRC and DOT regulations) has 
included grandfathering provisions. The 1985 and 1985 (as amended 1990) 
editions of Safety Series No. 6 contained provisions applicable to 
packages approved under the provisions of the 1967, 1973, and 1973 (as 
amended) editions of Safety Series No. 6. TS-R-1 includes provisions 
which apply to packages and special form radioactive material approved 
under the provisions of the 1973, 1973 (as amended), 1985, and 1985 (as 
amended 1990) editions of Safety Series No. 6.
    TS-R-1 grandfathering provisions (see TS-R-1, paragraphs 816 and 
817) are more restrictive than those previously in place in the 1985 
and 1985 (as amended 1990) editions of Safety Series No. 6. The primary 
impact of these two paragraphs is that packagings approved under the 
1967 edition of Safety Series No. 6 are no longer grandfathered, i.e., 
cannot be

[[Page 21405]]

used. The second impact is that fabrication of packagings designed and 
approved under Safety Series No. 6 1985 (as amended 1990) must be 
completed by a specified date. In regard to special form radioactive 
material, TS-R-1 paragraph 818 does not include provisions for special 
form radioactive material that was approved under the 1967 edition of 
Safety Series No. 6. Special form radioactive material that was shown 
to meet the provisions of the 1973, 1973 (as amended), 1985, and 1985 
(as amended 1990) editions of Safety Series No. 6 may continue to be 
used. However, special form radioactive material manufactured after 
December 31, 2003, must meet the requirements of TS-R-1. Within current 
NRC regulations, the provisions for approval of special form 
radioactive material are already consistent with TS-R-1.
    In TS-R-1, packages approved under Safety Series No. 6 1973 and 
1973 (as amended) can continue to be used through their design life, 
provided the following conditions are satisfied: multilateral approval 
is obtained for international shipment, applicable TS-R-1 QA 
requirements and A1 and A2 activity limits are 
met, and, if applicable, the additional requirements for air transport 
of fissile material are met. While existing packagings are still 
authorized for use, no new packagings can be fabricated to this design 
standard. Changes in the packaging design or content that significantly 
affect safety require that the package meet current requirements of TS-
R-1.
    TS-R-1 further states that those packages approved for use based on 
the 1985 or 1985 (as amended 1990) editions of Safety Series No. 6 may 
continue to be used until December 31, 2003, provided the following 
conditions are satisfied: TS-R-1 QA requirements and A1 and 
A2 activity limits are met and, if applicable, the 
additional requirements for air transport of fissile material are met. 
After December 31, 2003, use of these packages for foreign shipments 
may continue under the additional requirement of multilateral approval. 
Changes in the packaging design or content that significantly affect 
safety require that the package meet current requirements of TS-R-1. 
Additionally, new fabrication of this type packaging must not be 
started after December 31, 2006. After this date, subsequent package 
designs must meet TS-R-1 package approval requirements.
    Discussion. Industry representatives were concerned that IAEA is 
adopting a 2-year revision cycle to TS-R-1. From a design approval 
point of view, the regulatory requirements to be met may not be 
understood, and, as a new design requirement is approved, new revisions 
to the regulations could conceivably be developed. In other words, 
industry may always be playing catch up with the regulations.
    Previously, the IAEA standards permitted a package to be 
manufactured for two revision cycles of the IAEA standard. Because the 
IAEA standard was revised every 10 years, this equated to a 20-year 
period. However, IAEA is now changing to a 2-year revision cycle. 
Retaining the two-cycle provision would now equate to a 4-year 
allowable manufacturing period. This issue is under review by IAEA. 
Therefore, the NRC is proposing to specify in existing Sec. 71.13 when 
packages can no longer be manufactured or used, rather than using a 
``two-revision cycle'' approach.
    Additionally, a commenter expressed concern that beyond 2006, while 
packages could continue to be used under a valid CoC, no new packages 
could be manufactured based on any edition of Safety Series 6. 
Furthermore, all packages fabricated after December 31, 2006, would 
have to fully meet TS-R-1 requirements. The commenter stated that the 
licensing process for a package could be impacted. While NRC is aware 
and understands this concern, the proposed changes to Sec. 71.13 are 
adequate to address the potential limitation on fabrication and use.
    One commenter stated that the expense of designing and fabricating 
large Type B and spent fuel packages cannot be justified if the 
potential lifetime of the cask is limited to as short a period of time 
as 6 years. The commenter also believed that design and contents 
modifications should be allowed as specified in the current 
Sec. 71.13(c). Conversely, one commenter stated that a 2-year updating 
cycle would force safety considerations in cask design up front, rather 
than continuing the attitude that casks be used as long as possible.
    Another commenter urged NRC to include a grandfathering provision 
for continued transportation of packages, such as NRC-approved packages 
and DOT specification packages. The commenter explained that if NRC did 
not have a grandfathering provision, NRC would have to set aside 
hundreds of long-term disposal sites for the various Type B quantity 
containers currently in use at hospitals and research institutions.
    Several commenters believed that grandfathering would allow the NRC 
to maintain an adequate level of safety for package designs. Some 
commenters stated that existing packages (even older ones) were safe 
and durable, because these packages must be maintained in accordance 
with the QA regulations of Part 71. Another commenter added that under 
current regulations, NRC may immediately recall a certification if a 
particular package created a safety concern.
    One commenter voiced support for the proposal, assuming new 
regulations would continue to be more strict. Two commenters believed 
that while it is important for more stringent requirements to apply to 
all existing containers, relaxed provisions would effectively make 
newer containers less safe. In these instances, the commenters 
preferred that the older provisions remain in effect, instead of the 
newer, relaxed provisions. One commenter opposed grandfathering 
existing packages, and stated as a concern the unknown safety of older 
packages.
    One commenter believed that NRC should incorporate specific 
requirements into the grandfathering provision to effectively maintain 
a good package program. The commenter explained that manufacturers of 
CoC containers or packages should be allowed to show, by calculations 
or testing, that upgraded standards and TS-R-1 have been achieved.
    One commenter stated that the shorter cycle would put pressure on 
cask designers to make safety a more important design element.
    In response to the question about the type and magnitude of package 
design changes that should be allowed for grandfathered packages before 
recertification is required, two commenters stated that TS-R-1 allows 
for a phase out of manufacturing of any packages that are not certified 
to the 1996 version of TS-R-1 by December 31, 2006. The commenters 
added that this provides a window for the design, testing, and 
certification of new packages, the reevaluation of existing packages to 
the 1996 specification, or a request for special certification.
    The NRC recognizes that when the regulations change there is not 
necessarily an immediate need to discontinue use of packages that were 
approved under previous revisions of the regulations. Part 71 has 
included provisions that would allow previously-approved designs to be 
upgraded and to be evaluated to the newer regulatory standards. NRC 
believes that packages approved under the provisions of the 1967 
edition of Safety Series No. 6, and which have not been updated to 
later editions, may lack safety enhancements which have been included 
in the packages approved under the provisions of the 1973, 1973 (as 
amended), 1985 and 1985 (as amended 1990) editions of Safety Series No. 
6. Therefore, the NRC

[[Page 21406]]

believes that it is appropriate to begin a phased discontinuance of 
these earlier packages (1967-approved) to further improve transport 
safety.
    The following enhanced safety features have been included in NRC-
certified designs approved to these later standards. The NRC revised 10 
CFR Part 71 in 1983 for compatibility with the provisions of the 1973 
edition of Safety Series No. 6 to include:
    1. The introduction of the A1 and A2 system. 
Before the 1973 edition of Safety Series No. 6, the regulations were 
based on Transport Groups. The A1 and A2 system 
was intended to use a consistent safety basis for package contents 
based on radiological protection in transportation under normal and 
accident conditions.
    2. Standards for defining acceptable containment system 
performance. The 1973 edition of Safety Series No. 6 included for the 
first time activity limits for loss of radioactive contents from Type B 
packages under normal conditions of transport and under hypothetical 
accident conditions. The containment system performance requirements 
were tied to the A1 and A2 values, as described 
above.
    3. The immersion test for Type A fissile material packages. The 
1973 edition of Safety Series No. 6 required that the 15-meter (50-ft) 
water immersion test, previously required as a hypothetical accident 
test only for Type B packages, also be applied to fissile material 
packages. This immersion test is important in considering the degree of 
internal moderation (i.e., possible inleakage of water) in the 
criticality safety evaluation for fissile material packages in arrays.
    4. Maximum normal operating pressure (MNOP). The 1973 edition of 
Safety Series No. 6 added a revised definition of MNOP. The definition 
for MNOP was included in Part 71 and specifically excluded 
consideration of package venting and active cooling systems.
    5. Environmental test conditions. The 1973 edition of Safety Series 
No. 6 specified for the first time the high and low temperatures, 
pressures, and weights that should be considered when evaluating the 
package under normal and accident condition tests.
    6. Quality Assurance (QA) requirements. The requirements to apply 
QA to the design, fabrication, and use of transportation packages were 
proposed in Part 71 in 1973. Although the IAEA regulations did not 
adopt QA requirements until the 1985 edition of Safety Series No. 6, 
NRC regulations required QA controls before IAEA adopted these 
provisions. QA program requirements are only imposed on packages 
approved for use after 1979. Packages approved under the 1973 edition 
of Safety Series No. 6 include QA in their design and fabrication, 
whereas, with a few exceptions (such as spent fuel casks), packages 
approved under earlier editions do not include QA program requirements.
    The NRC draft RA indicates that adopting the grandfathering 
provisions for packagings approved under the 1985 editions of Safety 
Series No. 6 (known as ``-85'' packagings) and the associated 
expiration dates, is appropriate from a safety, regulatory, and cost 
perspective. From a regulatory standpoint, the proposed revisions would 
result in enhanced regulatory efficiency by bringing NRC's requirements 
in harmony with those contained in TS-R-1.
    NRC does not currently have sufficient information to quantify the 
economic impacts of adopting this provision. The estimated costs to 
industry are not quantifiable due to a lack of sufficient data. 
However, industry is expected to bear costs associated with the need to 
redesign existing packages, address the reduction in availability of 
packages, and determine the years of service expected from the original 
design. Should NRC receive comments providing detailed information on 
the potential economic impacts to industry, the draft RA would be 
revised accordingly.
    The proposed change would also result in implementation costs of 
approximately $3,500 to the NRC. The NRC would have to revise 
regulatory guides and NUREG-series documents to indicate which packages 
are covered by the ``grandfathering of older packages'' provision. 
Further, the proposed change could result in implementation and 
operation costs of approximately $1,000 to Agreement States if they 
adopt and implement parallel requirements. (The proposed change is not 
expected to affect implementation or operation costs of DOT.) Agreement 
States use regulatory guides and NUREG-series documents published by 
the NRC. Thus, Agreement States would only need to revise documents 
that they have specifically developed for their licensees (e.g., 
application materials). In terms of public health and safety, the 
existing and proposed requirements are believed to be equally 
protective. Thus, neither an increase nor a decrease in potential 
health and safety impacts is expected as a result of adopting the 
proposed administrative changes. Should the NRC become aware that a 
package or package design is unsafe, that package or design would be 
removed from service.
    NRC Proposed Position. NRC supports the update to grandfathering in 
TS-R-1 and is proposing to revise Part 71 to discontinue authorization 
to use packages approved under the provisions of the 1967 edition of 
Safety Series No. 6. Specifically, NRC is proposing to make 
modifications to existing Sec. 71.13 to phase out these types of 
packages. NRC realizes the impact this proposal may have on shipments 
using existing NRC-approved packages. Therefore, NRC proposes a 3-year 
transition period for the grandfathering provision on packages approved 
under the provisions of the 1967 edition of Safety Series No. 6. This 
period would provide industry the opportunity to phase out old packages 
and phase in new ones, or demonstrate that current requirements are 
met.
    For transitional arrangements for newer designs, NRC is proposing 
to incorporate into Sec. 71.13(c) the provisions for packagings 
approved under the 1985 editions of Safety Series No. 6 (known as ``-
85'' packagings) and the associated expiration dates. Additionally, 
paragraph (e) of Sec. 71.13 has been revised to specify the process by 
which previously-approved designs may be amended to include the ``-96'' 
designation.
    In summary, the following conditions would apply: (1) Packages 
approved under NRC standards that are compatible with the provisions of 
the 1967 edition of Safety Series No. 6 may no longer be fabricated, 
but may be used for a 3-year period after adoption of a final rule; (2) 
Packages approved under NRC standards that are compatible with the 
provisions of the 1973 or 1973 (as amended) editions of Safety Series 
No. 6 may no longer be fabricated; however, the proposed rule would not 
impose any restrictions on the use of these packagings; (3) Packages 
approved under NRC standards that are compatible with the provisions of 
the 1985 or 1985 (as amended 1990) editions of Safety Series No. 6, and 
designated as ``-85'' in the identification number, may not be 
fabricated after December 31, 2006, but may continue to be used; (4) 
Package designs approved under any pre-1996 IAEA standards (i.e., 
packages with a ``-85'' or earlier identification number) may be 
resubmitted to the NRC for review against the current standards. If the 
package design described in the resubmitted application meets the 
current standards, the NRC may issue a new CoC for that package design 
with a ``-96'' designation.
    Affected Sections. Sec. 71.13.

[[Page 21407]]

Issue 9. Changes to Various Definitions
    Background. The changes contemplated by NRC in this proposed 
rulemaking would require changes to various definitions in Sec. 71.4 to 
provide internal consistency and compatibility with TS-R-1. The terms 
must be clearly defined so that they can be used to accurately 
communicate requirements to licensees. By modifying existing 
definitions and adding new definitions, the licensee would benefit 
through more effective understanding of the requirements of Part 71.
    Discussion. Eight commenters submitted information on changes to 
various definitions in the proposed rule. One commenter stated that the 
definitions should be adopted to the extent the terms are used in the 
updated regulations. Another commenter urged NRC to be clear, 
consistent, and precise, particularly regarding the definitions of 
``rupture,'' ``collapse,'' ``buckling,'' and ``inleakage.'' Two other 
commenters stated that the TS-R-1 definition identifies the specific 
types of packaging allowed for Class 7, and unless DOT revises its 
regulations, there will be a domestic conflict. Therefore, these 
commenters do not recommend this change. The commenters added that NRC 
should consider definitions that explain the differences among 
``uniformly distributed,'' ``distributed throughout,'' and 
``homogeneous.''
    Another commenter stated that the existing regulation defines 
special form radioactive material that has been demonstrated to comply 
with specific tests. The commenter added that TS-R-1, paragraph 225, 
introduces the term ``low dispersible radioactive material,'' but fails 
to provide any guidance as to what characteristics qualify the 
material. Another commenter stated that the definition for ``low 
dispersible radioactive material'' should indicate that this does not 
refer to surface contamination, but rather activation of a solid 
material. This commenter also suggested adding the term ``sealed 
source'' to mean (for use of A1 values) encapsulated radioactive 
material that was designed and manufactured under a specific license 
and has been assigned a sealed source identification registry number.
    One commenter stated that the proposed definitions of ``confinement 
system'' and ``package'' are indistinguishable for packages intended to 
transport fissile material. The commenter urged NRC to use only one 
term or to clearly distinguish between the two definitions. The 
commenter added that if the definition of ``confinement system'' is 
added, the term ``competent authority'' must also be defined, and if 
the definition of ``package'' is incorporated, definitions of 
``excepted'' and ``industrial'' must be added. Another commenter stated 
that the confinement system definitions should be revised to include 
fuel assemblies, the PWR basket, and the shipping cask, because all 
three provide different levels and degrees of confinement.
    The NRC draft RA indicates that revising Part 71 to modify existing 
and add new definitions is appropriate from a safety, regulatory, and 
cost perspective. The proposed changes would provide greater internal 
consistency and compatibility with TS-R-1. By modifying existing 
definitions and adding new definitions, licensees would benefit through 
a more effective understanding of the requirements of Part 71.
    Specifically, industry will realize costs savings by benefitting 
from a more effective understanding of the requirements of Part 71. 
These costs savings are expected to be minimal, and are not 
quantifiable due to a lack of available data.
    The proposed changes would result in approximately $3,500 in 
implementation costs to the NRC. The NRC would have to revise 
regulatory guides and NUREG-series documents to include the new or 
revised definitions of Sec. 71.4. The proposed changes could affect 
implementation and operation costs of Agreement States because they 
would have to adopt the revision to the various definitions in 
Sec. 71.4. (The proposed change is not expected to affect 
implementation or operation costs of DOT.) Because Agreement States use 
regulatory guides and NUREG-series documents published by the NRC, they 
would only need to revise documents that they have developed 
specifically for their licensees.
    Additionally, as a means of improving use and understanding of Part 
71, the following existing definitions from Sec. 71.4 would be 
modified: A1, A2, and Low Specific Activity, 
specifically LSA-III. The definitions that are structured in Sec. 71.4 
are presented in italicized print as a means of distinguishing them 
from the corresponding text. The definition of LSA-III material would 
be modified to reference the testing provisions for LSA-III material 
found in Sec. 71.77. Other definitions (e.g., Special form radioactive 
material) reference requirements within Part 71 that must be followed.
    Lastly, within the Issues Paper, NRC posed the idea of adopting the 
following definitions from TS-R-1: Confinement System (TS-R-1, 
paragraph 209) and Quality Assurance (TS-R-1, paragraph 232). NRC is 
excluding the definition of Confinement system because it is included 
within the broader definition of Containment system. Further, NRC's use 
of Quality assurance is somewhat different from that of the IAEA, and 
NRC will retain the description of Quality assurance found in Subpart 
H.
    NRC Proposed Position. The NRC is proposing to adopt the TS-R-1 
definition of Criticality Safety Index (CSI). Additionally, the 
following definitions would be revised to improve their clarity: 
A1, A2, and LSA-III. Other changes to Sec. 71.4 
are proposed in separate issues.
    Affected Sections. Sec. 71.4.
Issue 10. Crush Test for Fissile Material Package Design
    Background. In TS-R-1, the crush test requirements have been 
broadened to apply to fissile material package designs (regardless of 
package activity). Previously, IAEA Safety Series No. 6 and Part 71 
have required the crush test for certain Type B packages. This 
broadened application was created in recognition that the crush 
environment was a potential accident force that should be protected 
against for both radiological safety purposes (packages containing more 
than 1,000 A2 in normal form) and criticality safety 
purposes (fissile material package design).
    Under requirements for packages containing fissile material, TS-R-
1, paragraph 682(b), requires tests specified in paragraphs 719-724 
followed by whichever of the following is the more limiting: (1) The 
drop test onto a bar as specified in paragraph 727(b) and either the 
crush test as indicated in paragraph 727(c) for packages having a mass 
not greater than 500 kg (1,100 lbs) and an overall density not greater 
than 1,000 kg/m\3\ (62.4 lbs/ft) based on external dimensions, or the 
9-meter (30-ft) drop test as defined in paragraph 727(a) for all other 
packages; or (2) the water immersion test as specified in paragraph 
729.
    Both the Safety Series No. 6, paragraph 548, and the current 
Sec. 71.73 require the crush test for packages having a mass not 
greater than 500 kg (1,100 lbs), an overall density not greater than 
1,000 kg/m\3\ (62.4 lbs/ft) based on external dimensions, and 
radioactive contents greater than 1,000 A2 not as special 
form radioactive material. Under TS-R-1, the criterion for radioactive 
contents greater than 1,000 A2 has been eliminated for 
packages containing fissile material. The 1,000 A2 criterion 
still applies to Type B packages and is also applied to the

[[Page 21408]]

IAEA newly created Type C package category.
    Discussion. Several commenters provided feedback regarding crush 
test requirements for packages containing fissile material. A number of 
commenters urged NRC to keep the current regulations requiring the 
crush test and the free drop test. One commenter stated that the crush 
test was especially useful for large packages. Another commenter 
supported the test and stated that U.S. transportation activities 
should be consistent with IAEA transportation regulations. Similarly, 
one commenter stated that the testing sequence as required in TS-R-1 
should be adopted to assure international uniformity. One commenter 
recommended removing the optional requirement of either a crush or a 
drop test, and replacing it with a requirement to conduct both tests.
    One commenter requested that NRC improve the realism associated 
with crush tests. The commenter stated that the crush test should be a 
physical test rather than using a computer model simulating a test. 
Additionally, the test should use full-scale packages that are loaded 
with nonradioactive materials to provide improved test reliability. 
This commenter stated that crush tests should be included for all 
package sizes, and the test parameters should be increased to reflect 
real-world conditions.
    A few commenters stated that the proposed requirement to use the 
free drop test or the crush test is problematic because the results of 
these tests are different and could require reanalysis of current 
packages.
    One commenter stated that elimination of the 1,000 A2 
activity limit, without providing for flexibility in test sequencing, 
would be an unfair and costly burden. The commenter stated that Part 71 
should be changed to conform to TS-R-1 in all aspects, or not be 
changed at all. Another commenter stated that the impact of the 
elimination of the 1,000 A2 activity limit for fissile 
material packages having a mass not greater than 500 kg (1,100 lbs), 
and overall density not greater than 1,000 kg/m\3\ (62.4 lbs/ft), based 
on external dimensions, is currently unknown. The commenter noted that 
shipping companies must use international standards established in TS-
R-1 to allow international trade. Another commenter supported the 
removal of the 1,000 A2 threshold for fissile packages on 
the grounds that A2 levels are intended as an index of 
radiological hazard rather than criticality potential, and it is 
inconsistent with TS-R-1.
    The NRC believes that full compliance with TS-R-1 requirements for 
fissile material packages would require changes to the hypothetical 
accident conditions test sequencing of Sec. 71.73 and would require 
performance of the 9-meter (30-ft) free drop test or the crush test, 
but not both, as presently required by Sec. 71.73. The TS-R-1 test 
requirements are essentially the same as those contained in Safety 
Series No. 6. In the previous NRC rulemaking for compatibility with 
Safety Series No. 6 (1985 edition), NRC staff addressed this difference 
in test requirements. In the June 8, 1988; 53 FR 21550, proposed rule, 
the NRC stated that: ``IAEA applies the crush test in place of the 9-
meter drop test for the lightweight packages specified. In the absence 
of experience using the crush test, and because the crush test and drop 
test evaluate different features of a package, NRC is requiring both 
the crush test and the 9-meter drop test for the lightweight 
packages.'' Further, in the September 28, 1995; 60 FR 50248, final 
rule, the NRC stated: ``NRC is requiring both the crush test and drop 
test, for lightweight packages, to ensure that the package response to 
both crush test and drop forces is within applicable limits.''
    The NRC draft RA indicates that revising Part 71 to adopt the TS-R-
1 requirements for a crush test for fissile material package design, 
while maintaining the current testing sequence, is appropriate from a 
safety, regulatory, and cost perspective. Not adopting the requirement 
would result in an inconsistency between Part 71 requirements and TS-R-
1, which could affect international shipments, and fissile material 
package designs would continue to not be evaluated for criticality 
safety against this potential accident condition. However, the NRC 
believes that further information on the impact of the TS-R-1 
requirement for fissile material package testing is required. Imposing 
the crush test requirement on fissile material package designs may 
impact the industry through costs imposed to demonstrate compliance and 
may lead to the redesign of packages. Under present Part 71 standards 
and Safety Series No. 6, the 1,000 A2 criterion, used to 
identify packages that must meet the crush test, essentially exempts 
all packages designed to contain uranium enriched to five percent or 
less (due to an unlimited A2 value). For fissile material 
package designs, this would only apply to designs for plutonium 
contents. However, if TS-R-1 is adopted, only the weight and density 
criteria would apply to fissile uranium material packages, and packages 
that were previously exempted because of the 1,000 A2 
criterion would now require crush testing. The potential impact on the 
industry is unknown due to a lack of data on the number of packages 
shipped under Sec. 71.55 where the 1,000 A2 value allowed 
exemption from crush testing. However, to demonstrate compliance with 
the new regulations, industry may incur additional costs. These 
potential costs may stem from package redesign but, due to the lack of 
available data, these costs are not quantifiable. NRC would bear 
approximately $74,000 in costs. These costs result from the need to 
prepare documents and conduct other activities (such as publishing 
notices of rulemakings, holding public hearings, and responding to 
public comments) as a result of the action.
    NRC Proposed Position. The NRC proposes to adopt the requirement 
for a crush test for fissile material packages, and eliminate the 1000 
A2 criterion for fissile material packages. However, because 
there is no new information that addresses concerns from the previous 
rulemaking regarding the difference in test requirements between Part 
71 and Safety Series No. 6, the NRC proposes not to change the testing 
sequence nor to change the drop and crush test requirements in this 
revision.
    Affected Sections. Sec. 71.73.
Issue 11. Fissile Material Package Design for Transport by Aircraft
    Background. TS-R-1 introduced new requirements for fissile material 
package designs that are intended to be transported aboard aircraft. 
TS-R-1 requires that shipped-by-air fissile material packages with 
quantities greater than excepted amounts (which would include all NRC-
certified fissile packages) be subjected to an additional criticality 
evaluation. Specifically, TS-R-1, paragraph 680, requires that packages 
must remain subcritical, assuming reflection by 20 centimeters (8 
inches) of water but no water inleakage (i.e., moderation) when 
subjected to the tests for Type C packages.\2\ The specification of no 
water ingress is given because the objective of this requirement is 
protection from criticality events resulting from mechanical 
rearrangement of the

[[Page 21409]]

geometry of the package (i.e., fast criticality). The provision also 
states that if a package takes credit for ``special features,'' this 
package can only be presented for air transport if it is shown that 
these features remain effective even under the Type C package test 
conditions followed by a water immersion test. ``Special features'' 
generally mean features that could prevent water inleakage (and 
therefore credit could be taken in criticality analyses) under the 
hypothetical accident conditions. Special features are permitted under 
current Sec. 71.55(c).
---------------------------------------------------------------------------

    \2\ The TS-R-1 imposition of Type C and LDM requirements (see 
Issue 6) was in recognition that severe aircraft accidents could 
result in forces exceeding those of the ``accident conditions of 
transport'' that are imposed on Type B and fissile package designs. 
Because the hypothetical accident conditions for Type B packages are 
the same as those applied to package designs for fissile material, 
there was also a need to consider how these more severe test 
conditions should be applied to fissile package designs transported 
by air.
---------------------------------------------------------------------------

    TS-R-1, paragraph 680, requirements for packages to be transported 
by air are in addition to the normal condition and accident tests that 
the package must already meet. Thus:
    Type A fissile package by air must:
    (A) Withstand normal conditions of transport with respect to 
release, shielding, and maintaining subcriticality (single package and 
5xN array \3\);
---------------------------------------------------------------------------

    \3\ N represents the maximum number of fissile material packages 
that can be shipped on a single conveyance.
---------------------------------------------------------------------------

    (B) Withstand accident condition tests with respect to maintaining 
subcriticality (single package and 2xN array); and
    (C) Comply with TS-R-1, paragraph 680, with respect to maintaining 
subcriticality (single package);
    Type B fissile package by air must:
    (A) Withstand normal conditions of transport and Type B tests with 
respect to release, shielding, and maintaining subcriticality (single 
package and 5xN array/normal and 2xN array/accident); and
    (B) Comply with TS-R-1, paragraph 680, with respect to maintaining 
subcriticality.
    There are no provisions in TS-R-1 for ``grandfathering'' (Issue 8) 
fissile material package designs, which will be transported by air. TS-
R-1, paragraphs 816 and 817, state that these packages are not allowed 
to be grandfathered. Consequently, all fissile package designs intended 
to be transported by aircraft would have to be evaluated before their 
use.
    Discussion. Five commenters provided information regarding our 
proposal of the TS-R-1 provisions for fissile material package design 
for transport by aircraft. One commenter expressed concern about the 
comprehensibility of the regulations for Type B or below quantities of 
fissile materials. The commenter was aware that the IAEA went through 
efforts to try to clarify the requirements, but asserted that the 
regulations need to be understood consistently by the people who 
approve package designs for transport of fissile materials by air. The 
commenter stated that this is a critical issue for industry because the 
International Civil Aviation Organization (ICAO) has adopted TS-R-1 in 
2001 and, therefore, shipments must meet the requirements in TS-R-1 for 
fissile materials. The commenter encouraged Federal agencies, including 
NRC and DOT, to push the concept of clarification of the rules and 
consider a streamlined approval process for designs of air transport of 
fissile material. Another commenter stated that TS-R-1 writers are 
working to develop a table that takes into consideration mass, 
enrichment, and moderation to define an acceptable limit for shipment 
by air.
    One commenter asked when and in what situations the transportation 
of fissile level material by air would be required.
    Two commenters supported the inclusion of these requirements as 
they are generally in parallel with those in place for surface mode 
accidents.
    The NRC draft RA indicates that adopting TS-R-1 paragraph 680 for 
criticality evaluation (only applicable to air transport) is reasonable 
from a safety, regulatory, and cost perspective. Adopting this change 
would provide the NRC with the regulatory framework for approving 
package designs that will be used internationally. Shippers will be 
required to meet these requirements even if the NRC does not adopt 
them, because the ICAO has adopted regulations consistent with TS-R-1 
on July 1, 2001. U.S. domestic air carriers require compliance with the 
ICAO regulations even for domestic shipments.
    These changes are expected to benefit industry by eliminating the 
need for two different package designs. The amount of these savings, 
however, are not quantifiable due to a lack of data.
    NRC Proposed Position. The NRC proposes to adopt TS-R-1, paragraph 
680, Criticality evaluation, in a new proposed Sec. 71.55(f) that only 
applies to air transport. Section 71.55 specifies the general package 
requirements for fissile materials, and the existing paragraphs of 
Sec. 71.55 are unchanged. Because (1) the NRC is deferring adoption of 
the Type C packaging tests (see Issue 6); (2) TS-R-1, paragraph 680, 
references the Type C tests; and (3) paragraph 680 applies to more than 
Type C packages, only the salient text would be inserted into 
Sec. 71.55(f), and would apply to domestic shipments.
    Affected Sections. Sec. 71.55.

D. NRC-Initiated Issues

Issue 12. Special Package Authorizations
    Background. The basic concept for radioactive material 
transportation is that radioactive contents are placed in an authorized 
container, or packaging, and then shipped. The packaging, together with 
its contents, is called the package. In general, the transportation 
regulations in TS-R-1, 10 CFR Part 71, and Title 49 are based on the 
shipment of radioactive contents in a separate, authorized packaging. 
There are a few exceptions, however. For example, TS-R-1 provides that 
the least radioactive of the Low Specific Activity materials (LSA-I) 
and Surface Contaminated Objects (SCO-I) may be shipped unpackaged, 
provided certain conditions are met. Title 49 permits shipment of LSA-I 
materials in bulk, where the conveyance (e.g., truck or freight 
container) serves as the packaging.
    In other cases involving larger quantities of radioactive material, 
the content to be shipped may itself be a container. A storage tank 
containing a radioactive residue is an example. It is not necessary for 
the shipper to place the tank within an authorized packaging, if the 
shipper demonstrates that the tank satisfies the requirements for the 
packaging. DOT and NRC have jointly provided guidance on such shipments 
(see ``Categorizing and Transporting Low Specific Activity Materials 
and Surface Contaminated Objects,'' NUREG-1608, RAMREG-003, July 1998).
    As older nuclear facilities are decommissioned, DOT and NRC are 
being asked to approve the shipment of large components, including 
reactor vessels and steam generators. These components may contain 
significant quantities of radioactive material, but they are so large 
that it is not practical to fabricate authorized packagings for them. 
Because these components were not contemplated when the regulations 
were developed, the regulations do not specifically address them.
    Basically, large components can be shipped under DOT regulations if 
the components meet the definition of Surface Contaminated Object (SCO) 
or Low Specific Activity (LSA) material (see 49 CFR 173.403 for SCO and 
LSA definitions). For example, steam generators that meet the SCO 
definition are exempt from Part 71 and are shipped under Title 49, 
following guidance provided in NRC Generic Letter 96-07 dated December 
5, 1996. This method has been applied to several shipments of steam 
generators and small reactor

[[Page 21410]]

vessels to the low level waste disposal facility at Barnwell, SC. NRC 
and DOT intend to continue employing this approach and method for steam 
generators and similar components that can be shipped under DOT 
regulations.
    Large components that exceed the SCO and LSA definitions are 
subject to Part 71. An example is the Trojan reactor vessel. By letter 
dated March 31, 1997, Portland General Electric Company (PGE) requested 
approval of the Trojan Reactor Vessel Package (TRVP) (including 
internals) for transport to the disposal facility operated by U.S. 
Ecology on the Hanford Nuclear Reservation near Richland, Washington. 
The TRVP contained approximately 74 PBq (2 million Ci) in the form of 
activated metal and 5.7 TBq (155 Ci) in the form of internal surface 
contamination, was filled with low-density concrete and weighed 
approximately 900 metric tons (1,000 tons). Normally, large curie 
contents are required to be shipped in a Type B packaging, but the TRVP 
was too large and massive to be shipped within another packaging.
    PGE acknowledged that the TRVP could not meet Type B regulations 
and applied for a Type B package CoC for the TRVP itself, either under 
Sec. 71.41(c), ``Demonstration of compliance,'' or Sec. 71.8, 
``Specific exemptions.'' Section 71.41(c) provides that ``Environmental 
and test conditions different from those specified in Secs. 71.71 and 
71.73 may be approved by the Commission if the controls proposed to be 
exercised by the shipper are demonstrated to be adequate to provide 
equivalent safety of the shipment.'' Section 71.41(c) has been used to 
accommodate minor deviations in test environments (e.g., initial 
temperatures), and was not intended to be used to establish new test 
conditions for Type B packages. The use of this provision in the Trojan 
case would essentially have resulted in establishing new (and less 
rigorous) Type B test conditions that the Trojan vessel could meet. A 
CoC for a Type B package could then have been issued for Trojan, but 
the level of performance reflected in that Certificate would have been 
significantly different from that in other Type B Certificates. NRC 
decided against using Sec. 71.41(c), and to use the Sec. 71.8 exemption 
provision--the only other option available.
    Section 71.8 provides that NRC may grant any exemption from the 
requirements of the regulations in Part 71 that it determines is 
authorized by law and will not endanger life or property nor the common 
defense and security. The exemption approach had three impacts on the 
TRVP review. First, the NRC's categorical exclusion from preparing an 
Environmental Assessment (EA) pursuant to the National Environmental 
Protection Act (NEPA) for package approvals (Sec. 51.22(c)(13)) does 
not apply to packages authorized under an exemption. Consequently, an 
EA of the proposed exemptions was required. Second, DOT's regulations 
that govern radioactive material shipments do not recognize packages 
approved via NRC exemption. PGE was therefore required to obtain an 
exemption from DOT regulations in 49 CFR Part 173 for the TRVP 
shipment. Third, use of the exemption option provided a mechanism for 
NRC to consider the operational and administrative controls, which were 
proposed by PGE to influence shipment risk factors. Considering the 
statements and representations contained in the application, as 
supplemented, and the conditions specified in the package approval, NRC 
concluded that the TRVP, as exempted, met the requirements of Part 71, 
and recommended that the Commission approve the exemptions and the TRVP 
shipment.
    Currently, no regulatory provisions exist in Part 71 for dealing 
with nonstandard packages, other than the exemption provisions and 
Sec. 71.41(c). The NRC's policy is to avoid the use of exemptions for 
recurring licensing actions. Therefore, as a lesson learned from the 
Trojan approval, the NRC staff identified large component package 
authorizations as an issue for consideration in this proposed rule.
    Discussion. Numerous comments were received on the special package 
approvals issue in response to the Issues Paper, from the public 
meetings, and from NRC's website. One of the commenters supported the 
idea of creating a system for providing special package approvals 
without using the existing exemption requirements. This commenter noted 
that his agency found it very useful to realize that there are packages 
or materials outside the current scope of NRC regulations that still 
need to be transported as they cannot stay where they are. The 
commenter agreed that it is appropriate to have a method to address 
these issues.
    A number of commenters did not support the development of a special 
package approvals regulation. These commenters believed the issue of 
special package approvals should be conducted on a case-by-case basis, 
using the current exemption process. One commenter noted that ``hot 
decommissioning'' and ``hot'' shipping introduce a new regimen, and 
therefore, the commenter believed that the only way for the NRC to 
proceed is with a case-by-case, very individual and specialized 
exemption or allowance, if at all. The commenter went on to say that 
the people who are on the first lines, the first responders and the 
emergency management coordinators at the local level, and the people 
who are in transport corridor communities have a right to information 
that a specialized process (i.e., an exemption process) would provide. 
The commenter stated that the concerns of the public who are in these 
transport corridor communities are not being given adequate weight in 
decision making, and the opportunities for discussion are too limited. 
Finally, this commenter stated that removing the exemption process for 
big, unusual shipments could set the stage for applying this concept to 
other types of materials to be exempted from testing and packaging 
requirements which the commenter believed would be a bad precedent.
    Two commenters expressed concern over the definition of a ``special 
large object.'' One commenter stated that if special provisions are 
added, then the term ``large'' must be defined with respect to both 
size and weight. Another commenter requested that NRC consider 
revisions to Part 71 to address large objects in general, that would 
include reactor vessels.
    Three commenters spoke to the issue of Type B quantities. The first 
commenter stated that there could be overlap between orphan sources and 
Type B quantities. This commenter recommended that Type B orphan 
sources be included in a separate rule from the special large packages. 
The second commenter would like to see collaboration between the NRC 
and DOT to address the possibility of initiating a program that would 
minimize package review costs of decommissioning Type B quantities of 
cobalt-60 and cesium-137. Two commenters stated that there have been 
cases where a Type B package has been damaged in a way that it will 
continue to secure and shield the sources, but does not meet compliance 
standards. The commenters noted that in these types of cases, a special 
arrangement certificate would be beneficial to allow transport of the 
damaged equipment for disposal.
    Several commenters did not believe that NRC's use of the shipment 
of the Trojan reactor vessel was an adequate basis for determining 
whether or not to remove the requirement for exemptions for special 
packages and replace it with other provisions. One commenter noted that 
because the Trojan vessel was

[[Page 21411]]

shipped by barge, a lot of the risk of exposure that would normally be 
present in other transport modes was removed (e.g., a truck being 
caught in traffic). This commenter also stated that moving to a risk-
informed decision making process for special package approvals may 
result in a situation where the public is ``informed to more risk while 
the industry is exposed to less regulation.'' Another commenter noted 
that if NRC is using the shipment of the Trojan reactor vessel as its 
baseline for determining whether to revise its regulations, care should 
be taken to limit the scope of this special approval to NRC's 
responsibilities and expertise. The commenter noted that as the Trojan 
approval process moved along, there was a difference of opinion as to 
the extent of NRC's evaluation of river and barging conditions, when in 
reality, these issues are the jurisdiction of the Coast Guard, and if 
the Coast Guard had approved the waterway and the conveyance, it should 
not be necessary for this information to be a part of an application to 
NRC subject to NRC review and approval. Other commenters disagreed. One 
commenter added that significant experience has already been gained in 
exempting the Trojan reactor vessel, a precedent has been established, 
and the possibility exists that the requirements placed on the shipment 
of the Trojan reactor vessel might have been more restrictive than 
might have been determined as necessary. Two commenters stated that the 
Trojan shipment review is a point of reference for the basis of other 
similar shipments, but that each case should still be assessed on its 
own merits.
    A number of commenters raised specific issues that NRC should 
consider when deciding whether to propose a special package approval 
process and how that process should be defined. Two commenters noted 
that the system has been defined as to how these materials should be 
moved and what kind of information needs to be provided to the 
regulators to move the materials. These commenters further noted that 
any change to Part 71, with respect to these special shipments, needs 
to be specific to those items that are going to be regulated under the 
MOU between the NRC and DOT. The two commenters added that the majority 
of those items that get moved are large components and would fall under 
the DOT's jurisdiction under the MOU. Thus, DOT would regulate items 
like steam generators and demineralizers and pressurizers, all of which 
are pieces and parts of reactors that are being decommissioned. NRC 
would regulate items like reactor pressure vessels (e.g., the Trojan 
reactor pressure vessel).
    One commenter did not support the adoption of an analog of the IAEA 
special arrangements provisions in Part 71. The commenter did not 
support the adoption of this type of provision in Part 71 because the 
IAEA special arrangements were specifically designed for movement 
internationally, whereas most of these items would be moved 
domestically.
    One commenter provided input on the specific issue of what 
additional determinations should be included in an application for a 
special package approval. The commenter noted that a precedent has 
already been established with the requirement that a transportation 
plan be provided with the exemption requests. The transportation plan 
contains safety features that would be substituted for the current 
codified requirements that would provide an equivalent order of safety, 
considerations of the entire safety system versus independent 
components of safety, emergency response plans, and risk-informed 
considerations.
    The NRC processing of one-time exemptions for nonstandard packages, 
such as the Trojan vessel, represents expenditure of considerable staff 
resources. Once the application for exemption is received, the staff 
spends a significant amount of time reviewing the application and 
preparing an EA. The Commission itself has been involved in the 
approval of these actions. Rather than exempting nonstandard packages 
from regulations, as was necessary for Trojan, the staff is proposing 
that regulatory requirements be added to Part 71 which would address 
nonstandard packages. These special packages are likely to increase in 
number as a result of future decommissioning activities.
    The NRC is proposing a regulatory mechanism to address large 
component shipments. In this regard, NRC has considered TS-R-1, 
paragraph 312, entitled Special Arrangement:

    Consignments for which conformity with the other provisions of 
these regulations is impracticable shall not be transported except 
under special arrangement. Provided the competent authority is 
satisfied that conformity with the other provisions of the 
regulations is impracticable and that the requisite standards of 
safety established by these regulations have been demonstrated 
through means alternative to the other provisions, the competent 
authority may approve special arrangement transport operations for 
single or a planned series of multiple consignments. The overall 
level of safety in transport shall at least be equivalent to that 
which would be provided if all the applicable requirements had been 
met. For international consignments of this type, multilateral 
approval shall be required.

    The Special Arrangement paragraph is intended to provide competent 
authorities (DOT in the U.S.) the authority to approve shipments that 
don't completely conform to the transportation safety standards, 
provided the overall level of safety established by the regulations is 
maintained. DOT consults with NRC regarding the approvals for shipment 
of packages containing larger quantities of radioactive material and/or 
fissile materials. NRC is proposing to add this provision to 
Sec. 71.41.
    The NRC draft RA indicates that adopting the special package 
authorization requirements proposed for incorporation into Part 71 is 
appropriate from a safety, regulatory, and cost perspective. The 
proposed action would result in enhanced regulatory efficiency by 
standardizing the requirements to provide greater regulatory certainty 
and clarity, and would ensure consistent treatment among licensees 
requesting authorization for shipment of special packages. This 
increase in regulatory efficiency, however, would depend in part on 
modifications to DOT's regulations to recognize NRC special package 
exemptions. Further, NRC experience in handling the one-time 
exemption(s) during the transition period would be used in crafting the 
new requirements. As a result, applications for one-time exemptions 
would be eliminated, resulting in savings in licensee staff resources 
and NRC staff resources. Because the new section is expected to be 
better streamlined for handling these nonstandard packages, 
considerable savings would be realized, both in NRC and licensee staff 
time. These expected NRC savings are estimated to be approximately 
$500,000. Special package shipments are likely to increase regardless 
of the outcome of this rulemaking, as a result of future 
decommissioning activities. The justification for authorizing special 
packages for shipment is a decreased risk of radiation exposure to the 
public and workers as opposed to the shipment alternatives. NRC 
believes that standardizing the method for reviewing these packages 
would provide adequate review without imposing unnecessary 
administrative burdens on NRC staff associated with the processing of 
exemption-based reviews.
    Industry may have costs associated with additional preparation of 
health and safety information for shipment of special packages. But, 
there may also be some inherent cost savings to industry

[[Page 21412]]

with respect to preparing health and safety information. On the balance 
between the costs anticipated with developing an application for NRC 
approval and the savings expected from using an established process, 
the net effect on industry is expected to be negligible.
    NRC Proposed Position. NRC proposes a special package authorization 
that would apply only in limited circumstances, and only to one-time 
shipments of large components. Further, any such special package 
authorization would be issued on a case-by-case basis, and would 
require the applicant to demonstrate that the proposed shipment would 
not endanger life or property nor the common defense and security, 
following the basic process used by applicants to obtain nonspecial 
package authorizations from NRC.
    NRC proposes to adopt a provision that is analogous to TS-R-1, 
paragraph 312, for Part 71 with respect to the approval of large 
component packages. The applicant would need to provide reasonable 
assurance that the special package, considering operational procedures 
and administrative controls employed during the shipment, would not 
encounter conditions beyond those for which it had been analyzed and 
demonstrated to provide protection. NRC would review applications for 
large component special package authorizations. Approval would be based 
on a staff determination that the applicant met the requirements of 
Subpart D. If approved, the NRC would issue a CoC or other approval 
(i.e., special package authorization letter).
    NRC would consult with DOT on making the determinations required to 
issue an NRC special package authorization. The efficiency of the NRC 
special package process, in part, depends on a modification by DOT of 
its regulations to recognize NRC special package authorizations, so 
that a DOT exemption would not be required for use of the NRC 
authorization. DOT is proposing this change in its companion TS-R-1 
compatibility rulemaking.
    Affected Sections. Sec. 71.41.
Issue 13. Expansion of Part 71 Quality Assurance Requirements to 
Certificate of Compliance (CoC) Holders
    Background. The Commission recently issued a final rule to expand 
the QA provisions of Part 72, Subpart G, to specifically include 
certificate holders and applicants for a CoC (see 64 FR 56114; October 
15, 1999). In development of the proposed rule for Part 72, the NRC 
staff submitted a rulemaking plan to the Commission in SECY-97-214.\4\ 
In a Staff Requirements Memorandum (SRM) to SECY-97-214, the Commission 
approved the staff's rulemaking plan and directed the staff to also 
consider whether conforming changes to the QA regulations in Part 71 
would be necessary because of the existence of dual-purpose cask 
designs. In a memorandum from the Executive Director for Operations to 
the Commission, dated December 3, 1997, the NRC staff indicated that 
expansion of the Part 71 QA provisions to include certificate holders 
and applicants for a CoC would be made as part of the rulemaking to 
conform Part 71 to IAEA Standard TS-R-1. Furthermore, in the final rule 
expanding QA regulations in Part 72, Subpart G, the Commission did not 
include contractors or subcontractors (e.g., fabricators) within the 
scope of the revised Part 72, Subpart G. The Commission took this 
action in response to comments on the associated proposed rule. In the 
response to Comments 3 and 9 in the final Part 72 rule, the Commission 
indicated that Part 72 licensees, certificate holders, and applicants 
for a CoC are responsible for assuring that their contractors and 
subcontractors (e.g., fabricators) are implementing adequate QA 
programs. Similarly, Part 71 licensees, certificate holders, and 
applicants for a CoC are responsible under Sec. 71.115 for assuring 
that their contractors and subcontractors (e.g., fabricators) are 
implementing adequate QA programs.
---------------------------------------------------------------------------

    \4\ SECY-97-214, ``Changes to 10 CFR Part 72, Expand 
Applicability to Include Certificate Holders and Applicants and 
Their Contractors and Subcontractors,'' dated September 24, 1997.
---------------------------------------------------------------------------

    Under Part 71, the NRC reviews and approves applications for Type B 
and fissile material packages for the transport of radioactive 
material. The NRC's approval of a package is documented in a CoC. 
Applicants for a CoC are currently required by Sec. 71.37 to describe 
their QA program for the design, fabrication, assembly, testing, 
maintenance, repair, modification, and use of the proposed package. 
Further, existing Sec. 71.101(a) describes QA requirements that apply 
to design, purchase, fabrication, handling, shipping, storing, 
cleaning, assembly, inspection, testing, operation, maintenance, 
repair, and modification of components of packagings that are important 
to safety. Type B packages are intended to transport radioactive 
material that contains quantities of radionuclides greater than the 
A1 or A2 limits for each radionuclide (see 
Appendix A to Part 71 for examples of A1 or A2 
limits). Fissile material packages are intended to transport fissile 
material in quantities greater than the Part 71, Subpart C, general 
license limits for fissile material (e.g., existing Secs. 71.18, 71.20, 
71.22, and 71.24).
    Although CoCs are legally binding documents, certificate holders or 
applicants for a CoC and their contractors and subcontractors have not 
clearly been brought into the scope of Part 71 requirements. This is 
because the terms ``certificate holder'' and ``applicant for a 
certificate of compliance'' do not appear in Part 71, Subpart H; 
rather, Subpart H only mentions ``licensee'' in these regulations. 
Consequently, the NRC has not had a clear basis to cite certificate 
holders and applicants for a CoC for violations of Part 71 requirements 
in the same way it has licensees.
    The NRC Enforcement Policy \5\ and its implementing program was 
established to support the NRC's overall safety mission in protecting 
public health and safety and the environment. Consistent with this 
purpose, enforcement actions are used as a deterrent to emphasize the 
importance of compliance with requirements and to encourage prompt 
identification and comprehensive correction of the violations. 
Enforcement sanctions consist of Notices of Violation (NOVs), civil 
penalties, and orders of various types. In addition to formal 
enforcement actions, the NRC also uses related administrative actions 
such as Notices of Nonconformance (NONs), Confirmatory Action Letters, 
and Demands for Information to supplement its enforcement program. The 
NRC expects licensees, certificate holders, and applicants for a CoC to 
adhere to any obligations and commitments that result from these 
actions and would not hesitate to issue appropriate orders to ensure 
that these obligations and commitments are met. The nature and extent 
of the enforcement action are intended to reflect the seriousness of 
the violation involved. An NOV is a written notice setting forth one or 
more violations of a legally binding requirement.
---------------------------------------------------------------------------

    \5\ NUREG-1600, ``General Statement of Policy and Procedures for 
NRC Enforcement Actions,'' dated May 2000.
---------------------------------------------------------------------------

    When the NRC has identified a failure to comply with Part 71 QA 
requirements by certificate holders or applicants for a CoC, it has 
issued an NON rather than an NOV. Although an NON and an NOV appear to 
be similar, the Commission prefers the issuance of an NOV because: (1) 
The issuance of an NOV effectively conveys to both the person violating 
the requirement and the public that a violation of a legally binding

[[Page 21413]]

requirement has occurred; (2) the use of graduated severity levels 
associated with an NOV allows the NRC to effectively convey to both the 
person violating the requirement and the public a clearer perspective 
on the safety and regulatory significance of the violation; and (3) 
violation of a regulation reflects the NRC's conclusion that potential 
risk to public health and safety could exist. Therefore, the NRC 
believes that limiting the available enforcement sanctions to 
administrative actions is insufficient to address the performance 
problems observed in industry.
    Discussion. Sixteen commenters provided comments regarding the 
possible expansion of QA requirements to holders of, and applicants 
for, a CoC. Of these, three supported expanding the QA requirements. 
Two commenters stated that the cask design and fabricating industry 
should be allowed flexibility to make design changes to the casks that 
would not impact safety. One of the commenters stated that cask 
designers and fabricators should be held responsible as are parties on 
the nuclear power reactor side.
    Four commenters did not support the overall proposed change to 
expand the QA requirements of Part 71. One commenter stated that it is 
the responsibility of the purchaser, user, or licensee of the cask or 
shipping container to ensure the container's QA, and therefore, NRC 
already has enforcement authority over that particular container. Two 
commenters stated that extending the responsibility to the fabricator 
or certificate holder would encourage fabricators to get out of 
business because of the regulatory and paper burden of the proposed 
provision. Another commenter stated that there is confusion between 
what is in the current regulations and what is in the proposed 
regulations. Another commenter stated that NRC could be regulating 
packages for which NRC is not responsible under the MOU between the NRC 
and the DOT. A commenter stated that NRC currently has adequate QA 
control on the Part 71 packages under Subpart H. The commenters did not 
believe that issuing an NOV instead of an NON would result in 
additional compliance.
    Several commenters noted the need for consistency in the QA 
provisions between Parts 71 and 72, which should be maintained for dual 
purpose casks used for storage and transportation of spent nuclear fuel 
and high-level radioactive waste. Additionally, one commenter noted 
that a distinction has never been established between Part 71 and Part 
72 packages used to transport/store spent fuel and the Part 71 packages 
used to transport sealed radioactive sources. The commenter suggested 
that ``Part 50 reactor licensees be specifically exempted from 
participation in nuclear power specific QA activities.''
    Representatives of DOT and DOE questioned whether this provision 
would apply to Type A packages. The NRC intends that this proposed 
change would apply only to NRC certificate holders and applicants for a 
CoC and only for package designs that are regulated by NRC (e.g., Type 
B or fissile packages).
    The principal changes to Subpart H would involve adding the terms 
``certificate holder'' and ``applicant for a CoC'' to indicate that 
these persons are also covered by the section, although in some cases, 
only ``certificate holder'' would be added, because an applicant for a 
CoC would not be expected to accomplish these specific activities. 
Additional conforming changes would be made to various sections in Part 
71 to ensure greater consistency between Part 71 and Part 72.
    The NRC draft RA indicates that expanding the QA provisions of Part 
71, Subpart H, to certificate holders and applicants for a CoC is 
appropriate from a safety, regulatory, and cost perspective. Adopting 
these requirements would ensure that the regulatory scheme of Part 71 
would remain more consistent with other NRC regulations in that 
certificate holders and applicants for a CoC would be responsible for 
the behavior of their contractors and subcontractors. Also, because 
this action would be limited to certificate holders and applicants for 
a CoC, it may not be as likely to be challenged as an expansion of NRC 
authority. Inclusion of certificate holders and applicants for a CoC 
would make it possible for NRC to issue NOVs and orders, if 
appropriate, for violation to the regulatory requirements; this would 
allow the NRC to conduct its business of protecting public health and 
safety more efficiently and effectively. This proposed rule would not 
authorize the NRC to issue civil penalties to Part 71 certificate 
holders or applicants for a CoC who are found to be in violation of 
regulatory requirements. Alternatively, contractors and subcontractors 
of licensees, certificate holders, and applicants do have 
responsibility for safety, and omitting them from Part 71 would 
continue the present difficulty that NRC has encountered in reaching 
these persons with its enforcement tools. Certificate holders and 
applicants for a CoC would incur costs associated with understanding 
and implementing the new regulations, as well as in preparing and 
submitting reports similar to those described in SECY 99-174. SECY 99-
174 states that ``Additional requirements for recordkeeping and 
reporting for certificate holders are needed to include records 
required to be kept as a condition of the CoC. This will provide an 
enforcement basis equivalent to the recordkeeping and reporting 
regulations for licensees.'' These costs are estimated to be 
approximately $239,000 per year for the certificate holders and 
applicants for a CoC. NRC would incur costs associated with monitoring 
certificate holders and applicants for a CoC and maintaining and 
reviewing the records for certificate holder submittals. These costs 
are estimated to be approximately $48,000 per year. By specifically 
listing certificate holders and applicants for a CoC in Part 71, 
inspection deficiencies noted by NRC might result in an NOV. This 
authority would allow NRC to issue orders or take other enforcement 
actions (except civil penalties) necessary to ensure that certificate 
holders and applicants for a CoC comply with Part 71 requirements, 
similar to NRC enforcement actions in other program areas. However, 
this benefit is difficult to quantify and is estimated to be small.
    The NRC is proposing to expand the QA provisions of Part 71, 
Subpart H, to specifically include certificate holders and applicants 
for a CoC. This expansion is necessary to enhance NRC's ability to 
enforce nonconformance by the certificate holders and applicants for a 
CoC. The NRC is also proposing to add a new section (Sec. 71.9) on 
employee protection to Part 71. Currently, regulations on employee 
protection are contained in the individual parts under which the NRC 
issued a specific license. Consequently, this regulation was not deemed 
necessary for a Part 71 general licensee. However, the equivalent 
requirement for certificate holders or applicants for a CoC does not 
exist. The NRC believes that employee protection regulations should be 
added for the employees of certificate holders and applicants for a CoC 
to provide greater regulatory equivalency between Part 71 licensees and 
certificate holders. Therefore, the NRC would add a requirement on 
employee protection to Part 71.
    NRC Proposed Position. The NRC is proposing to expand the QA 
provisions of Part 71, Subpart H, to specifically include certificate 
holders and applicants for a CoC.
    In addition to the changes to Subpart H, conforming changes would 
also be made to: Sec. 71.0, ``Purpose and scope''; Sec. 71.1, 
``Communications and records'';

[[Page 21414]]

Sec. 71.6, ``Information collection requirements: OMB approval''; 
Sec. 71.7, ``Completeness and accuracy of information''; Sec. 71.91, 
``Records''; Sec. 71.93, ``Inspection and tests''; and Sec. 71.100, 
``Criminal penalties.'' Additionally, Sec. 71.11 would be redesignated 
as Sec. 71.8; and a new Sec. 71.9, ``Employee protection,'' would be 
added.
    Affected Sections. Secs. 71.0, 71.1, 71.6, 71.7, 71.8 , 71.9, 
71.91, 71.93, 71.100, and 71.101 through 71.137.
Issue 14. Adoption of American Society of Mechanical Engineers (ASME) 
Code
    Background. NRC considered the adoption of the American Society of 
Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (B&PV) 
Code, Section III, Division 3, for two reasons. First, previous NRC 
inspections at vendor and fabricator shops (for fabrication of spent 
fuel storage canisters and transportation casks) identified quality 
control (QC) and QA problems. Some of these problems would have been 
prevented with improved QA programs, and may have been prevented had 
fabrication occurred under more prescriptive requirements such as the 
ASME Code requirements. Second, Public Law 104-113, ``National 
Technology Transfer and Advancement Act,'' enacted in 1996, requires 
that Federal agencies use, as appropriate, consensus standards (e.g., 
the ASME B&PV Code), except when there are justified reasons for not 
doing so.
    Currently, no ASME Code requirements exist in Part 71 for 
fabrication/construction of spent fuel transportation packages.
    Discussion. NRC received numerous comments regarding the adoption 
of the ASME Code. Four commenters stated they favored adoption of the 
ASME Code. One commenter favored using ASME codes for all components 
used in the containment boundary of all products that are used in 
transportation and storage of radioactive materials. This commenter 
also supported an explanatory guideline in the ASME Code that speaks to 
the subject of categorization of materials, whereby all manufacturers 
are using the same criteria. Another commenter stated that using ASME 
standards would improve current problems with casks and the current 
lack of QA. One commenter stated that some benefits of a third party 
authorized nuclear inspector (ANI) would accrue to industry. These 
benefits are that common standards would decrease complexity and 
interpretation, lower cost, and increase safety.
    Eight commenters stated concerns or disapproval of the adoption of 
the ASME Code. One commenter was concerned with the adoption of the 
guidelines before a full review of the effects on transportation. 
Another commenter stated concern over adopting voluntary standards into 
regulations. Specifically, this concern was directed at the 
inconsistency between industry standards and regulations. Similarly, 
another commenter noted that changes within ASME might occur quickly, 
and it would be difficult to follow these changes. One commenter 
recommended that incorporation of the ASME Code by reference is the 
appropriate regulatory mechanism, following the precedent set by 
Sec. 50.55(a) for the ASME Code, Section III, Division 1. Several 
commenters recommended that NRC place industry standards in regulatory 
guides, which would allow for simpler updating, recognize that other 
methods of demonstrating compliance are available, and satisfy the 
Congressional mandate to consider the use of consensus standards. One 
commenter stated a concern about the enforceability of the standard if 
it is not placed in the regulations. Conversely, another commenter 
noted that the regulatory burden is significantly increased when 
voluntary standards are changed to regulations, and compliance may not 
always be practical or accomplished.
    Other commenters were concerned about the widespread impact of the 
adoption. One commenter stated that there is no technical justification 
for adoption of the ASME Code, and it would have significant adverse 
impact on the ability of the U.S. Navy to refuel and defuel the U.S. 
nuclear powered warships. Another commenter stated that overseas market 
impacts need to be considered in the rulemaking. Another commenter 
stated that when an applicant commits to certain standards in his or 
her safety requirements during the license approval process, it becomes 
a license condition, and NRC can enforce it.
    One commenter stated that if the ASME Code is adopted, the 
development of it and the information involved must be publicly 
available. Two commenters specifically asked if the proposed change 
applies to all packages, dual-purpose spent fuel packages, or to all 
CoC holders. Another commenter questioned how, or whether, the 
requirement will change if the industry standard changes in the future.
    During the early period of spent fuel storage and transportation 
cask fabrication, NRC inspection staff consistently identified QC and 
QA problems at the vendor/fabricator facilities. At that time, NRC 
believed that these problems might have been prevented had fabrication 
occurred under ASME Code requirements. Therefore, there was an impetus 
to place consideration of the ASME Code requirements in the Part 71 
rulemaking. However, since then, due to increased attention by the NRC 
and industry, the overall frequency and significance of QA and QC 
problems at fabricators and vendors have decreased.
    With respect to conformance to Public Law 104-113, the ASME issued 
a consensus standard in May 1997, entitled: ``Containment Systems and 
Transport Packages for Spent Fuel and High Level Radioactive Waste,'' 
ASME B&PV Code, Section III, Division 3. The ASME Code requires the 
presence of an ANI during construction to ensure that the ASME Code 
requirements are met, and the stamping of components (i.e., the 
transportation cask's containment) constructed to the ASME Code. NRC 
staff participated, and continues to participate, in the ASME 
subcommittee that developed the ASME Code requirements. It is the NRC 
staff's understanding, through participation in the subcommittee, that 
the ASME Code document is undergoing extensive review and modification 
and that a major revision will be issued. Therefore, NRC staff believes 
that inclusion of the ASME Code in Part 71 is not appropriate at this 
time.
    Public Law 104-113 requires that Federal agencies use consensus 
standards in lieu of government-unique standards, if this use is not 
impractical or inconsistent with other existing laws. Because a major 
revision to the ASME Code is forthcoming and because the changes in 
that revision are not yet available for staff and stakeholder review, 
the NRC staff considers it an imprudent use of NRC and stakeholder 
resources to initiate rulemaking on the current ASME Code revision only 
to have the ASME Code requirements change during the Part 71 
rulemaking. After the ASME Code revision is issued, the NRC staff can 
then consider its incorporation through the rulemaking process, or 
consider adopting and accepting the ASME Code as an acceptable method 
for complying with NRC requirements through endorsement in regulatory 
guidance.
    The NRC draft RA indicates that not adopting the ASME Code 
requirements in Part 71 is appropriate from a safety, regulatory, and 
cost perspective. While NRC resources would be conserved by not 
adopting the ASME Code, the proposed action would retain the current 
status. However, the proposed

[[Page 21415]]

action would result in no benefits or negative impacts on industry.
    After consideration of the public comments and the NRC recently 
learning of the extensive review and revision of the ASME Code, the 
staff recommends not to incorporate the ASME Code, Section III, 
Division 3, requirements into Part 71. However, adoption of the ASME 
Code into Part 71 will be considered by the NRC staff in a future 
rulemaking or guidance document.
    NRC Proposed Position. The NRC staff recommends not incorporating 
the ASME Code, Section III, Division 3 requirements into Part 71.
    Affected Sections. None (not adopted).
Issue 15. Change Authority for Dual-Purpose Package Certificate Holders
    Background: The Commission recently approved a final rule to expand 
the provisions of Sec. 72.48, ``Changes, Tests, and Experiments,'' to 
include Part 72 certificate holders (64 FR 53582; October 4, 1999). 
Part 72 certificate holders are allowed under the amended Sec. 72.48 to 
make certain changes to a spent fuel storage cask's design or 
procedures used with the storage cask and to conduct tests and 
experiments, without prior NRC review and approval. Part 71 does not 
contain any similar provisions to permit a certificate holder to change 
the design of a Part 71 transportation package, without prior NRC 
review and approval. The NRC has issued separate CoCs under Parts 71 
and 72 for dual-purpose spent fuel casks and transportation packages 
(i.e., a container intended for both the storage and transportation of 
spent fuel). This has created the situation where an entity holding 
both a Part 71 and Part 72 CoC would be allowed under Part 72 to make 
certain changes to the design of a dual-purpose cask, e.g., changes 
that affected a component or design feature that has a storage 
function, without obtaining prior NRC approval. However, the same 
entity would not be allowed under Part 71 to make changes to the design 
of this same dual-purpose cask (package), e.g., changes that affect the 
same component or design feature, if that component or feature also has 
a transportation function, without obtaining prior NRC approval, even 
when the same physical component and change is involved (i.e., the 
change involves a component that has both storage and transportation 
functions).
    In SECY-99-130 \6\ and SECY-99-054,\7\ NRC indicated that comments 
had been received on the Sec. 72.48 proposed rule (63 FR 56098; October 
21, 1998) that requested similar authority be created in Part 71, 
particularly with respect to dual-purpose casks. In SECY-99-054, NRC 
staff recommended that an authority similar to Sec. 72.48 be created 
for spent fuel transportation packages intended for domestic use only. 
NRC staff also recommended that this authority be limited to Parts 50 
and 72 licensees shipping spent fuel and the Part 71 certificate 
holder. NRC indicated that providing change authority under Part 71 
would be addressed in the current rulemaking. The Commission directed 
the staff to implement recommendations contained in SECY-99-130 and 
SECY-99-054, in an SRM dated June 22, 1999.
---------------------------------------------------------------------------

    \6\ SECY-99-130; May 12, 1999, ``Final Rule--Revisions to 
Requirements of 10 CFR Parts 50 and 72 Concerning Changes, Tests, 
and Experiments.''
    \7\ SECY-99-054; February 22, 1999, ``Plans for Final Rule--
Revisions to Requirements of 10 CFR Parts 50, 52, and 72 Concerning 
Changes, Tests, and Experiments.''
---------------------------------------------------------------------------

    NRC also identified other supporting changes to Part 71 that would 
be required to ensure consistency with the process contained in 
Sec. 72.48. These changes include: (1) the use of common terminology 
such as ``changes to the cask design, as described in the final safety 
analysis report'' (FSAR); (2) a process for requesting amendments to a 
CoC; (3) periodic updates by certificate holders to the FSAR for a 
transportation package to ensure that an accurate ``licensing'' basis 
is available when future proposed changes are evaluated; and (4) a 
requirement that licensees possess a copy of the FSAR as well as the 
CoC before making a shipment.
    NRC believes that the current IAEA standard TS-R-1 does not contain 
any equivalent provisions for changing a transportation package's 
design, without prior review by the agency that certified the design. 
NRC is the reviewing agency for Type B and fissile material package 
approvals. Therefore, any application of ``change authority'' to Part 
71 CoCs would only apply to packages intended for the domestic 
transport of spent fuel.
    Discussion. The NRC has received 48 public comments on this issue 
in response to the issue paper, public meetings, and the website. 
Industry representatives and certain members of the public support the 
issue. Public interest organizations, State representatives, and other 
members of the public generally oppose the issue. The DOE also opposes 
this issue. Groups in favor of this issue pointed to similar provisions 
in Parts 50 and 72 where such changes have been safely made. Groups 
opposed to this issue believe that all changes to a transport package's 
design should be submitted to the NRC for prior review and approval. 
These commenters believed this is necessary because transportation 
packages are on the public roadways and railways, hence the public 
believes there is more immediate and greater exposure to the 
radioactive contents of the package in an accident. The following is a 
more detailed description of these comments.
    Seven commenters supported the effort to expand the provisions 
contained in Sec. 72.48 to include Part 71 certificate holders. Two 
commenters also requested that NRC expand the authority for all 
packages, not just dual-purpose spent nuclear fuel packages.
    Three commenters requested that NRC be consistent and revoke the 
change, test, and experiment authority for Part 72 certificate holders. 
One commenter opposed allowing the ability to make any changes to casks 
without prior NRC approval. Similarly, one commenter sought assurance 
that NRC would continue to be able to monitor industry performance 
(i.e., maintain regulatory oversight capability), and be able to undo 
or revise changes or force amendments when necessary.
    One commenter, opposed to the expansion of authority, referenced a 
Government Accounting Office (GAO) report that highlighted problems 
with transportation casks fabricated by Westinghouse, claiming that 20 
out of 40 casks had been found to be defective. Another commenter was 
opposed to any action, such as moving to performance-or risk-based 
management, that would increase the level and type of public risk.
    Another commenter stated that he does not support allowing change 
authority because the definition of ``minimal'' has historically been 
ill-defined. This commenter also expressed his belief that Issue 15 
(change authorization issue), as currently proposed, would not result 
in Part 71 conforming with TS-R-1. The commenter cited as evidence the 
text in the Issues Paper that states, ``the current IAEA standard ST-1 
does not contain any equivalent provisions for changing a 
transportation package's design, without prior review by the competent 
authority.''
    Most commenters expressed interest in receiving additional 
information from NRC about what changes might be allowable, and 
clarification that these allowable changes would only be for activities 
not important to safety (e.g., switching to nonreactive paints). One

[[Page 21416]]

commenter also suggested that NRC and DOT be careful in determining 
allowable, nonsafety changes because with the effort to lengthen the 
certificate revalidation cycle, it is conceivable that these changes 
would just be rolled into the new certification without review. This 
commenter also questioned how NRC plans to address the issue of 
conformity with other nations' package requirements and certificates.
    NRC believes that the capability to make minor changes to a 
transportation package is similar to the capability to make minor 
changes to a reactor facility, to a spent fuel storage facility, or a 
spent fuel storage cask design. The Commission has recently issued a 
final rule which authorized Part 72 certificate holders to make minor 
changes to a spent fuel storage cask's design. Therefore, NRC believes 
that extending this authority to Part 71 packages is consistent with 
previous Commission actions.
    The current regulatory structure of Part 71 requires that all 
design changes to a transportation package, which would change the CoC 
or included drawings, be submitted to the NRC for prior review and 
approval. However, a package user (i.e., a Part 71 general licensee) is 
not currently required to obtain a copy of the safety analysis report 
(SAR) and understand it before shipping radioactive material. Rather, 
the licensee is only required to obtain a copy of the CoC and any 
referenced documents, determine that the package is properly configured 
for shipment (i.e., meets the requirements of Secs. 71.85 and 71.87), 
determine that the intended radioactive contents are within the 
conditions of the CoC, implement any procedure required by the CoC, and 
accomplish these activities under an NRC-approved QA program (in 
accordance with Part 71, Subpart H). Consequently, a licensee is not 
required to understand the technical bases of the Part 71 regulations 
on normal conditions of transport, hypothetical accident conditions, 
and criticality control (i.e., Secs. 71.71, 71.73, and 71.55, 
respectively), before the licensee can use the package to transport 
radioactive material. Therefore, NRC staff believes that a significant 
increase in burden would be imposed on licensees to understand these 
technical bases, if they were permitted to make changes under a 
``change authority'' regulation.
    NRC also notes that Part 71 does not contain some of the regulatory 
foundations which support the recent revision to Sec. 72.48. For 
example, under Sec. 72.48, a licensee is required to evaluate proposed 
changes to the cask design against the FSAR (as updated), and to 
periodically incorporate these changes into the FSAR to ensure that an 
accurate licensing basis is maintained for use in evaluating future 
proposed changes. Additionally, a Part 71 licensee need not own the 
package it is using to transport radioactive material. Instead, the 
licensee is considered a ``registered user'' of the package. This 
second circumstance, when coupled with a Part 71 change authority, 
might create a situation in which one licensee could make an authorized 
change to a package, without prior NRC approval, transfer that package 
to another registered user, without forwarding all change summaries to 
the next user, who would then be unable to verify or recognize that the 
package is in conformance with the CoC (i.e., acceptable for use under 
the requirements of Subpart G (e.g., Sec. 71.87)).
    The design drawings for a transportation package are directly 
incorporated by reference into the Part 71 CoC, whereas the design 
drawings for a spent fuel storage cask are contained in the FSAR. While 
changes to a design (as described in the FSAR) are permitted, changes 
to the CoC (or any drawings incorporated into the CoC by reference) 
would not be permitted. As a consequence, these referenced drawings 
limit the population of potential changes that a licensee or 
certificate holder could make under a Part 71 change authority 
equivalent to Sec. 72.48.
    Based upon review of the potential impacts, NRC believes that 
adding the necessary regulatory requirements (i.e., foundations) to 
Part 71 to support a change authority equivalent to Sec. 72.48 would 
unnecessarily increase the burden on all licensees without providing a 
corresponding benefit. Providing this change authority would also 
increase the complexity of the Part 71 regulations.
    The NRC believes the issue of inconsistent change authority between 
Parts 71 and 72 for a dual-purpose spent fuel package should be 
resolved. Performance of Parts 50 and 72 licensees and the Part 76 
certificate holder in implementing the change processes of Parts 50, 
72, and 76 has demonstrated that these types of changes can be made 
safely, without prior NRC approval. However, NRC staff also believes 
that the scope of this authority should be limited to dual-purpose 
packages, rather than all NRC-certified spent fuel packages, and 
limited to only the certificate holders.
    Accordingly, the NRC staff considers the best approach in resolving 
these conflicts is through the use of a parallel regulatory structure 
in Part 71. While the NRC staff would retain the current process for 
existing transportation packages, a new process for approving dual-
purpose transportation packages would be added to Part 71. Authority to 
make changes to a dual-purpose package design would be provided, and 
new requirements on the issuance and review of an SAR would also be 
provided. These new regulations would only apply to Type B(DP) dual-
purpose packages intended for the domestic transportation and storage 
of spent fuel. Because IAEA standard TS-R-1 does not contain any 
provisions to permit a certificate holder to make changes to the design 
of a package without prior review and approval by the ``competent 
authority'' that issued the certificate, a Type B(DP) package could not 
be approved for international use.
    To provide a clear distinction between these new and existing 
packages, the new packages would be classified as Type B(DP), would 
have a unique ``B(DP)'' identifier, and for reasons discussed below, 
these packages would not be required to meet TS-R-1 standards and could 
not be used in international transport. For a Type B(DP) package, 
requirements on submitting an FSAR, periodically updating the FSAR, 
applying for an amendment to the CoC, and changing the design of the 
dual-purpose package, without prior NRC approval, would be consolidated 
in a new Subpart I to Part 71. To provide greater consistency between 
the Parts 71 and 72 CoCs, the NRC staff would use the same 20-year term 
for both CoCs and would synchronize the CoCs' expiration dates. 
Further, the NRC staff would use the same 20-year term for a QA program 
approval to design or fabricate a Type B(DP) package.
    Additionally, a new general license (Sec. 71.18) would be added to 
Subpart C that would require a licensee shipping spent fuel in a Type 
B(DP) package to have both a copy of the CoC and the current updated 
FSAR before making the shipment. Licensees would not be authorized 
under this proposed rule to make changes to a Type B(DP) package's 
design by themselves, but would be required to obtain certificate 
holder (i.e., the package designer) review and approval of the proposed 
change. Further, should the evaluation of the proposed change indicate 
that prior NRC approval is required, then only the certificate holder 
would be authorized to submit an application to the NRC to amend the 
CoC.
    NRC believes that approval of proposed changes to the design of a 
Type B(DP) package, or submitting a request to modify a package's 
design,

[[Page 21417]]

should be restricted to the certificate holder. As described above, 
licensees have not previously been required to understand the design 
bases for a transportation package or the technical bases of the Part 
71 regulations.
    The NRC believes that the new parallel structure provides a choice 
to applicants desiring to obtain transportation certification for a 
spent fuel storage and transportation package. This proposed structure 
(in Subpart I) would not restrict an applicant's right to obtain a CoC 
for a spent fuel transportation package under the existing requirements 
in Subpart D. Applicants can weigh the costs and benefits associated 
with each approach against the needs of its customers and determine 
which approach is better. Consequently, the NRC believes the new 
parallel structure is voluntary and does not impose a backfit.
    Additional conforming changes would be made to Sec. 71.0 to include 
Type B(DP) packages within the scope of Part 71; to Sec. 71.4 to add a 
definition for Certificate of compliance, Type B(DP) packages, and 
Structures, systems, and components important to safety; to Sec. 71.6 
to reflect the new recordkeeping and reporting requirements created by 
the addition of new Subpart I (required under the Paperwork Reduction 
Act); to add a new Sec. 71.10 to provide for public availability of 
applications; to Sec. 71.51 to exclude Type B(DP) packages; and to 
Sec. 71.100 to indicate which of these new sections (i.e., Sec. 71.18 
and Subpart I) would be subject to criminal penalties.
    The NRC draft RA indicates that the proposed expansion of Part 71 
to include a new Sec. 71.175, ``Changes, tests, and experiments,'' to 
include Part 71 certificate holders is reasonable from a regulatory, 
cost, and safety perspective. As noted, however, NRC has very limited 
data from which to draw this conclusion. The NRC believes that not 
adopting these provisions may be awkward and appears to result in a 
regulatory inconsistency. Specifically, this inconsistency appears in 
situations where a certificate holder for a dual-purpose cask design 
could not modify the design of a component that had both storage and 
transport functions without prior NRC approval, irrespective of the 
certificate holder's authority under Sec. 72.48 to modify the design of 
a storage cask. While the adoption of this change would not be 
consistent with the requirements in TS-R-1, NRC believes the benefits 
to be gained by allowing Part 50 and Part 72 licensees and the Part 71 
certificate holder to revise the cask design for a dual-purpose cask 
outweigh the potential impacts of this inconsistency. Further, these 
impacts would be offset by restricting this authority to packages 
intended for domestic shipments only. Preliminary estimates indicate 
that NRC costs would decline slightly by adopting this change, because 
NRC would not have to review as many license amendments each year. This 
cost savings was determined to be negligible in the Sec. 72.48 
regulatory analysis, and would be offset by the agency having to adopt 
new document controls to handle the ``minimal change'' submission 
required every 2 years for licensees making ``minimal changes.'' For 
the 350 recordkeeping licensees listed in the Part 71 Supporting 
Statement, professional judgment was used to assume that, in any given 
year, 50 percent of licensees will perform a ``minimal change'' as 
described in Sec. 72.48 over a 2-year period. Submittals under 
Sec. 72.48 are required every 2 years; therefore, approximately 88 
submittals are expected per year. The cost savings of reporting 
``minimal changes'' versus preparing license amendments is estimated at 
approximately $2.4 million per year. The 350 licensees would incur a 
one-time recordkeeping cost of approximately $2.3 million the first 
year this change is implemented.
    NRC Proposed Position. The NRC proposes to add a new type of 
package (dual-purpose) to Part 71 [i.e., Type B(DP)]. Type B(DP) 
transportation packages would be certified for the storage of spent 
fuel under Part 72 and for transportation of spent fuel under Part 71. 
Type B(DP) packages would be restricted to use in domestic commerce. 
Requirements on the submission, review, amendment, and issuance of a 
CoC for a Type B(DP) package would be contained in a new Subpart I to 
Part 71. A new general license providing for the use of a Type B(DP) 
package would be added to Subpart C (Sec. 71.18). Certificate holders 
for Type B(DP) packages would also be required to submit, and 
periodically update, an FSAR describing the package's design. 
Additionally, only the certificate holder for a Type B(DP) package 
would be allowed under Subpart I to make changes to the package's 
design.
    Additionally, conforming changes would be made to Secs. 71.0, 71.4, 
71.6, 71.10, 71.17, and 71.100
    Affected Sections. Secs. 71.0, 71.4, 71.6, 71.10, 71.17, 71.18, 
71.100, and 71.151 through 71.177.
Issue 16. Fissile Material Exemptions and General License Provisions
    Background. The NRC published an emergency final rule amending its 
regulations on shipments of small quantities of fissile material (62 FR 
5907; February 10, 1997). This rule revised the regulations on fissile 
exemptions in Sec. 71.53 and the fissile general licenses in 
Secs. 71.18 and 71.22. The NRC determined that good cause existed, 
under Section 553(b)(B) of the Administrative Procedure Act (APA) (5 
U.S.C. 553(b)(B)), to publish this final rule without notice and 
opportunity for public comment. Further, the NRC also determined that 
good cause existed, under Section 553(d)(3) of the APA (5 U.S.C. 
553(d)(3)), to make this final rule immediately effective. 
Notwithstanding the final status of the rule, the NRC provided for a 
30-day public comment period. The NRC subsequently published in the 
Federal Register (64 FR 57769; October 27, 1999) a response to the 
comments received on the emergency final rule and a request for 
information on any unintended economic impacts caused by the emergency 
final rule.
    The NRC issued this emergency final rule in response to a 
regulatory defect in the fissile exemption regulation in Sec. 71.53 
which was identified by an NRC licensee. The licensee was evaluating a 
proposed shipment of a special fissile material and moderator mixture 
(beryllium oxide mixed with a low concentration of high-enriched 
uranium). The licensee concluded that while Sec. 71.53 was applicable 
to the proposed shipment, applying the requirements of Sec. 71.53 
could, in certain circumstances, result in an inadequate level of 
criticality safety (i.e., an accidental nuclear criticality was 
possible in certain unique circumstances).\8\
---------------------------------------------------------------------------

    \8\ For transportation purposes, ``nuclear criticality'' means a 
condition in which an uncontrolled, self-sustaining, and neutron-
multiplying fission chain reaction occurs. ``Nuclear criticality'' 
is generally a concern when sufficient concentrations and masses of 
fissile material and neutron moderating material exist together in a 
favorable configuration. Neutron moderating material cannot achieve 
criticality by itself in any concentration or configuration. 
However, it can enhance the ability of fissile material to achieve 
criticality by slowing down neutrons or reflecting neutrons.
---------------------------------------------------------------------------

    The NRC staff confirmed the licensee's analysis that this beryllium 
oxide and high-enriched uranium mixture created the potential for 
inadequate criticality safety during transportation. An added factor in 
the urgency of the situation was that under the NRC regulations in 
Secs. 71.18, 71.20, 71.22, 71.24, and 71.53, these types of fissile 
material shipments could be made without prior approval of NRC. For 
many years, NRC allowed these shipments of small quantities of fissile 
material based on NRC's understanding

[[Page 21418]]

of the level of risk involved with these shipments, as well as 
industry's historic transportation practices. This experience base had 
led NRC (and its predecessor, the Atomic Energy Commission (AEC)) to 
conclude that shipments made under the fissile exemption provisions of 
Part 71 typically required minimal regulatory oversight (i.e., NRC 
considered these types of shipments to be inherently safe).\9\
---------------------------------------------------------------------------

    \9\ The NRC's regulations in part 71 ensure protection of public 
health and safety by requiring that Type AF, B, or BF packages used 
for transportation of large quantities of radioactive materials be 
approved by the NRC. This approval is based upon the NRC's review of 
applications which contain an evaluation of the package's response 
to a specific set of rigorous tests to simulate both normal 
conditions of transport (NCT) and hypothetical accident conditions 
(HAC). However, certain types of packages are exempted from the 
testing and NRC prior approval; these are fissile material packages 
that either contain exempt quantities (Sec. 71.53), or are shipped 
under the general license provisions of Secs. 71.18, 71.20, 71.22, 
or 71.24.
---------------------------------------------------------------------------

    All public comments on the emergency final rule supported the need 
for limits on special moderators (i.e., moderators with low neutron-
absorption properties such as beryllium, graphite, and deuterium). 
However, the commenters stated that the restrictions were far too 
limiting (to the point that some inherently safe packages were excluded 
from the fissile exemption) and could lead to undue cost burdens with 
no benefit to safety. In addition, the commenters believed that the 
consignment mass limits set to deter undue accumulation of fissile mass 
would be extremely costly. Therefore, the commenters recommended that 
further rulemaking was necessary to resolve these excessive 
restrictions. Based on the public comments on the emergency final rule, 
NRC staff contracted with Oak Ridge National Laboratory (ORNL) to 
review the fissile material exemptions and general license provisions, 
study the regulatory and technical bases associated with these 
regulations, and perform criticality model calculations for different 
mixtures of fissile materials and moderators. The results of the ORNL 
study were documented in NUREG/CR-5342,\10\ and NRC published a notice 
of the availability of this document in the Federal Register (63 FR 
44477; August 19, 1998). The ORNL study confirmed that the emergency 
final rule was needed to provide safe transportation of packages with 
special moderators that are shipped under the general license and 
fissile material exemptions, but the regulations may be excessive for 
shipments where water moderation is the only concern. The ORNL study 
recommended that NRC revise Part 71.
---------------------------------------------------------------------------

    \10\ NUREG/CR-5342, ``Assessment and Recommendations for 
Fissile-Material Packaging Exemptions and General Licenses Within 10 
CFR Part 71,'' July 1998.
---------------------------------------------------------------------------

    Subsequently, NRC published a Federal Register notice that 
responded to public comments on the emergency final rule and requested 
additional information on the cost impact of the emergency final rule 
from the public, industry, and DOE (64 FR 57769; October 27, 1999). The 
Commission requested this cost impact information because the NRC staff 
was not successful in obtaining this information. Specifically, NRC 
requested information on the cost of shipments made under the fissile 
material exemptions and general license provisions of Part 71 before 
the publication of the emergency final rule, and those costs and/or 
changes in costs resulting from implementation of the emergency rule. 
One commenter agreed with the NRC approach, but stated that, ``the 
limits for those materials containing no special moderators can and 
should be increased, hopefully back to their pre-emergency rule 
levels.''
    As part of NUREG/CR-5342, ORNL performed computer model 
calculations of keff (k-effective) for various combinations 
of fissile material and moderating material, including beryllium, 
carbon, deuterium, silicon-dioxide, and water, to verify the accuracy 
of current minimum critical mass values. These minimum critical mass 
values were then applied to the regulatory structure contained in Part 
71, and revised mass limits for both the general license and exemption 
provisions to Part 71 were determined. Also, ORNL researched the 
historical bases for the fissile material exemption and general license 
regulations in Part 71 and discussed the impact of the emergency final 
rule's restrictions on NRC licensees. ORNL concluded that the 
restrictions imposed by the emergency final rule were necessary to 
address concerns relative to uncontrolled accumulation of exempt 
packages (and thus fissile mass) in a shipment and the potential for 
inadequate safety margin for exempt packages with large quantities of 
special moderators.
    Based on its new keff calculations, ORNL suggested that: 
(1) The mass limits in the general license and exemption provisions 
could be safely increased and thereby provide greater flexibility to 
licensees shipping fissile radioactive material; and (2) additional 
revisions to Part 71 were appropriate to provide increased 
clarification and simplification of the regulations. Copies of NUREG/
CR-5342 may be obtained by writing to the Superintendent of Documents, 
U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402-
9328. Copies are also available from the National Technical Information 
Service, 5285 Port Royal Road, Springfield, VA 22161-0002. A copy is 
also available for inspection and copying, for a fee, at the NRC Public 
Document Room in the NRC Headquarters at One White Flint North, Room O-
1F23, 11555 Rockville Pike, Rockville, MD 20852-2738.
    Discussion. The NRC has received public comments on this issue in 
response to the Issues Paper, public meetings, and the workshop. 
Industry representatives, public interest organizations, Agreement 
States, and members of the public supported the issue. None of the 
comments presented new issues from those previously presented in 
response to the emergency final rule or the Commission's request for 
additional cost information.
    Addressing the emergency final rule, one commenter agreed with the 
necessity for the rule, but stated that there are issues yet to be 
resolved for water moderated shipments. In comparison, another 
commenter took issue with our stated goal and NRC's methods. This 
commenter believed that if NRC adopts these provisions, then NRC will 
be unable to conform with TS-R-1. The commenter cited as evidence a 
statement in the issues paper, ``IAEA standard ST-1 (nee TS-R-1) 
contains language on fissile exemptions and restrictions on the use of 
special moderators. However, ST-1 does not currently contain provisions 
on general licenses for shipment of fissile material.''
    Similarly, one commenter raised the importance of coordinating 
regulatory actions on fissile material exemptions with the 
international community. The commenter noted the international 
community's interest in fissile material exemptions and encouraged NRC 
to listen to its international counterparts at the next IAEA meeting; 
the commenter's goal being to ensure that NRC is not out of step with 
the rest of the world (i.e., fissile material exempt in the U.S. is not 
exempt elsewhere, and vice-versa).
    One commenter raised questions concerning specific recommendations 
in NUREG/CR-5342. The commenter was concerned in how recommendations 3 
and 4 would introduce unnecessary complexity and noted that this 
concern vanishes if the TS-R-1 definitions for regulated material are 
adopted. The commenter also stated that recommendation 17 could 
seemingly eliminate the fissile excepted category,

[[Page 21419]]

which is something the commenter did not want to see occur. If such a 
change is necessary, the commenter requested that the NRC instead 
revise the excepted package's definition to reduce the amount of 
fissile material present and ensure that 10 CFR 71.53 and 49 CFR 
173.453 are consistent with TS-R-1 (i.e., with respect to upper limits 
on a package's fissile material, as well as the total amount of fissile 
material in a fissile exempt consignment).
    The current restrictions on fissile exempt and general license 
shipments under Secs. 71.53, and 71.18 through 71.24, respectively, are 
burdensome for a large number of shipments that actually contain no 
special moderating materials (i.e., packages that are shipped with 
water considered as the potential moderating material). This problem 
was clearly expressed in public comments on the emergency final rule. 
Another regulatory problem is that the current fissile exempt and 
general license provisions are cumbersome and outdated; this was one of 
the main conclusions of the ORNL study. Therefore, the NRC would 
update, simplify, and streamline these sections of Part 71 to eliminate 
regulatory confusion.
    The proposed revisions in Table 16-1 are based on public comments 
received on the February 10, 1997, emergency final rule, on the 
subsequent Commission's direction in SRM-SECY-99-200 regarding the 
unintended economic impact of that emergency final rule, and on the 
latest public comments received on the July 2000 Issues Paper. 
Altogether, ORNL suggested 17 changes to the Part 71 regulations in 
NUREG/CR-5342. A summary of these changes and the NRC's assessment and 
recommendation are contained in Table 16-1. NUREG/CR-5342 contains a 
more detailed discussion of the proposed changes listed in Table 16-1 
and ORNL's supporting calculations.

      Table 16-1.--Summary of Recommended Changes in NUREG/CR-5342
------------------------------------------------------------------------
           Description of issue               NRC staff recommendation
------------------------------------------------------------------------
Issue 16-1: Definitions for                 Disagree. These changes are
 ``consignment,'' ``consignor,'' and         not necessary with the use
 ``shipper'' should be provided to reduce    of mass ratio limits and a
 confusion between regulations in 49 CFR     criticality safety index
 Part 173 and 10 CFR Part 71.                when combined with the
                                             current requirement in Sec.
                                              71.59.
Issue 16-2: Plutonium-238 should be         Agree.
 removed from the definition of ``fissile
 material,'' because 238Pu is only
 fissionable, not fissile.
Issue 16-3: The exemption for radioactive   Disagree. The existing
 material in Sec.  71.10(a) should be        exception to the exemption
 revised to exclude fissile material.        in paragraph (b) would be
 ORNL's concern was that a large quantity    maintained (i.e., the
 of a low-concentration fissile material     reference to the fissile
 could pose a criticality safety concern.    exemption in new Sec.
 The revised keff calculations indicate      71.15). However, no change
 that a 43 Bq/g (1.16  x  10-3 Ci/  would be made to paragraph
 g) limit for fissile material (235U)        (a) because the values in
 would be necessary. However, other          Table A-2 are less than 43
 fissile nuclides have higher limits (e.g,   Bq/g (1.16  x  10-3 Ci/g) for        m>Ci/g) or the fissile
 \233\U or 66,000 Bq/g (1.784 Ci/   nuclides have criticality
 g) for 241Pu) or the Appendix A, new        limits which would be
 Table A-2, values are only 10 Bq/g (2.7)    higher than the exempt
 x  10-4 uCi/g) (e.g., 239Pu).               concentration limits of
                                             Table A-2.
Issue 16-4: The exemption for radioactive   Agree.
 material in existing Sec.  71.10 should
 be revised to require shipment in an
 acceptable package as required by
 existing Sec.  71.11 to improve safety.
Issue 16-5: Section 71.53 should be         Agree.
 relocated from Subpart E--Package
 Approval Standards, to Subpart B--
 Exemptions, to provide greater
 consistency in Part 71. (Note: Sec.
 71.53 would also be redesignated as Sec.
 71.15.).
Issue 16-6: The NRC or DOT should keep a    Disagree. The licensee's
 database of shipments made under the        burden in keeping and
 fissile exemption or general licenses.      reporting these records is
 Section 71.97 should be revised to          not commensurate with the
 require licensees to keep these records     safety risk for fissile
 and report this information.                exemption shipments.
Issue 16-7: The provisions for plutonium-   Agree.
 beryllium (Pu-Be) shipments should be
 removed from the four general licenses of
 existing Secs.  71.18, 71.20, 71.22, and
 71.24 and consolidated in a new general
 license. The mass limits for Pu-Be
 shipments should be reduced, because the
 revised keff calculations indicate
 potential safety problems exist with the
 current limits.
Issue 16-8: The general licenses of         Agree.
 existing Secs.  71.18, 71.20, 71.22, and
 71.24 should be consolidated into one
 general license to simplify the
 regulations and consistently apply the
 criticality safety index (CSI).
Issue 16-9: The distinction between         Agree.
 quantities of 235U that can be shipped in
 a uniform distribution and nonuniform
 distribution should be eliminated from
 the general licenses. The bounding
 nonuniform quantities should be used to
 simplify compliance with the rule.
Issue 16-10: Restrictions on the            Agree.
 quantities of Be, C, and D2O to less than
 0.1% should be removed for the general
 licenses. A maximum of 500g of Be, C, and
 D2O per package should be imposed to
 preclude the potential for these
 materials to be effective as reflector
 materials.
Issue 16-11: A separate mass control or     Agree.
 restriction for moderators having a
 hydrogen density greater than water
 should be retained for general licenses.
 For mixtures of moderators, lower mass
 limits should be imposed if more than 15%
 of the moderating material has a
 moderating effectiveness greater than the
 hydrogen density of water. Use of a 15%
 mixture limit would reduce confusion when
 mixtures of moderators are present in a
 shipment.
Issue 16-12: Package mass limits for        Agree. Also, minimum package
 general licenses may be increased to        requirements should be
 reflect results of new analyses and still   established. However,
 maintain equivalence of safety as           imposing Sec.  71.43
 provided for requirements certified         requirements would be
 packages.                                   excessive for the
                                             commensurate risk from
                                             these shipments. Instead,
                                             the DOT Type A package
                                             requirements should be
                                             used.
Issue 16-13: Package mass limits for        Agree.
 general licenses should be revised to
 reflect the new keff calculations. These
 mass limits can be safely increased.

[[Page 21420]]

 
Issue 16-14: The mass-limit based
 exemption in existing Sec.  71.53(a)
 should be changed to a mass-ratio based
 approach. In contrast to concentration-
 based approaches with consignment limits
 that are now in use in the fissile
 exemptions, the mass-ratio approach
 should provide a simpler, more cost-
 effective approach to preventing the
 formation of system configurations having
 inadequate subcritical margins as a
 result of transport scenarios (Secs.
 71.71 and 71.73).
Issue 16-15: If a mass-ratio approach is    Agree.
 used, the restrictions on Be, C, and D2O
 in existing Sec.  71.53(a), (c), and (d)
 should be removed.
Issue 16-16: The exemption for uranyl       Agree in part. Minimum
 nitrate solutions in Sec.  71.53(c)         package requirements should
 should include a packaging requirement      be established. However,
 from existing Sec.  71.43.                  Sec.  71.43 is excessive
                                             for the commensurate risk
                                             from these shipments. The
                                             DOT Type A package
                                             requirements should be
                                             used.
Issue 16-17: The exemption for uranium      Agree.
 enriched to less than 1 wt % 235U in
 existing Sec.  71.53(b) should be
 modified to remove the homogeneity
 requirements and lattice prevention
 requirement. Instead, retain the 0.1% Be,
 C, and D2O limit because of the
 difficulty in defining and applying
 ``homogenous'' and ``lattice
 arrangement'' restrictions.
------------------------------------------------------------------------

    In addition to the recommendations contained in NUREG/CR-5342, the 
Commission directed the NRC staff, in SRM-M970122B on SECY-96-268, to 
issue additional guidance in instances where fissile materials may be 
mixed in the same shipping container with different moderators (i.e., 
materials of differing moderator effectiveness). Therefore, the NRC 
would add a note to Table 71-1 in existing Sec. 71.22 to use reduced 
mass limits if more than 15 percent of the moderating materials in a 
package have a moderating effectiveness greater than the average 
hydrogen density of H2O (see Issue 16-11 in Table 16-1 
above).
    The NRC believes these changes would provide greater flexibility in 
the shipment of fissile material under the fissile exemption and 
general license regulations. The NRC would revise these requirements 
using a risk-informed approach, and address the burden and 
excessiveness issues raised in the public comments on the emergency 
final rule. The NRC would use a graduated regulatory approach in 
establishing requirements for the shipment of fissile material. The 
graduated approach would involve three tiers of regulations consisting 
of: (1) The fissile material exemptions with low fissile mass limits 
and minimal requirements (i.e., the new Sec. 71.15); (2) the fissile 
general licenses with higher mass limits and packaging and QA 
requirements (i.e., the new Secs. 71.22 and 71.23); and (3) the Type 
AF, BF, B(U)F, or B(M)F fissile material packages with large mass 
limits that require prior NRC approval of the package design (i.e., the 
existing Sec. 71.55). The NRC believes this approach would establish a 
risk-informed framework by imposing progressively stricter requirements 
as the quantity of fissile material being shipped increases (i.e., the 
criticality hazard increases). In accomplishing this risk-informed 
approach, some mass limits in the general licenses would increase, and 
others would decrease. These changes would reflect the new 
Keff calculations in NUREG/CR-5342. To counterbalance the 
increases in mass limits in the general licenses, requirements would be 
added on the use of a Type A package, a CSI, and an NRC-approved QA 
program.
    While the NRC is proposing to adopt the use of the CSI for general 
licensed fissile packages, the NRC is proposing to retain the current 
per package (CSI) limit of 10, rather than raising the per package 
limit to 50 (see Issue 5). TS-R-1 does not address the issue of fissile 
general licenses, so no compatibility issues arise with retention of 
the current NRC per package limit of 10. NRC staff believes that 
because reduced regulatory oversight is imposed on fissile general 
license shipments (e.g., the package standards of Secs. 71.71 and 
71.73, fissile package standards of Sec. 71.55, and fissile array 
standards of Sec. 71.59 are not imposed for fissile general license 
shipments), retention of the current per package limit of 10 is 
appropriate. Furthermore, retention of the current per package limit of 
10 would not impose a new burden on licensees; rather, licensees 
shipping fissile material under the general license provisions of 
Secs. 71.22 and 71.23 would not be permitted to take advantage of the 
relaxation of the per package CSI limit from 10 to 50 that would be 
permitted for Types A(F) and B(F) package shipments.
    Overall, the NRC would amend Part 71 as follows: (1) Revise 
Sec. 71.10, ``Exemption for low level material,'' to exclude fissile 
material, also redesignate Sec. 71.10 as 71.14; (2) redesignate 
Sec. 71.53 as Sec. 71.15, ``Exemption from classification as fissile 
material,'' and revise the fissile exemptions; (3) consolidate the 
existing four general licenses in existing Secs. 71.18, 71.20, 71.22, 
and 71.24 into one general license in new Sec. 71.22, revise the mass 
limits, and add Type A package, CSI, and QA requirements; and (4) 
consolidate the existing general license requirements for plutonium-
beryllium sealed sources, which are contained in existing Secs. 71.18 
and 71.22 into one general license in new Sec. 71.23 and revise the 
mass limits. Additionally, conforming changes would be made to 
Sec. 71.4, ``Definitions'' and Sec. 71.100, ``Criminal penalties.''
    The NRC draft RA indicates that incorporating revisions to the 
fissile material exemption and general license provisions in Part 71 is 
appropriate from a safety, regulatory, and cost perspective. As stated 
earlier, there is a shortage of data on the fissile material general 
license and exempt shipments; consequently, the NRC was not successful 
in obtaining data to quantify the economic impact which would result 
from adopting some or all of the 17 recommendations in NUREG/CR-5342. 
The impact of these amendments on the licensees and the NRC would be 
both positive and negative, depending on the specific recommendation. 
Recommendations 1, 2, and 5 would enhance regulatory efficiency due to 
the increase in clarity of the NRC regulations. Recommendations 3, 4, 
6, 9, and 12 would increase costs to licensees. Recommendations 7, 8, 
10, 13, 14, and 15 would eliminate the potential for criticality 
accidents, which would, in turn, yield environmental and

[[Page 21421]]

public health and safety benefits. Finally, recommendations 11, 16, and 
17 would result in savings to licensees.
    NRC Proposed Position. The NRC proposes revisions to the fissile 
material exemptions and the general license provisions in Part 71.
    Affected Sections. Secs. 71.4, 71.10, 71.11, 71.18, 71.20, 71.22, 
71.24, 71.53, 71.59, and 71.100.
Issue 17. Double Containment of Plutonium (PRM-71-12)
    Background: In 1974, the AEC issued a final rule which imposed 
special requirements on the shipment of plutonium (39 FR 20960; June 
17, 1974). These requirements are located in Sec. 71.63 and apply to 
shipments of radioactive material containing quantities of plutonium in 
excess of 0.74 TBq (20 curies). Section 71.63 contains two principal 
requirements. First, the plutonium contents of the package must be in 
solid form (Sec. 71.63(a)). Second, the packaging containing the 
plutonium must provide a separate inner containment (i.e., the ``double 
containment'' requirement) (Sec. 71.63(b)). In addition, the AEC 
specifically excluded from the double containment requirement of 
Sec. 71.63(b) plutonium in the form of reactor fuel elements, metal or 
metal alloys, and other plutonium-bearing solids that the Commission 
(AEC or NRC) may determine, on a case-by-case basis, do not require 
double containment. This regulation remained essentially unchanged from 
1974 until 1998, when vitrified high-level waste in sealed canisters 
was added to the list of exempt forms of plutonium in Sec. 71.63(b) (63 
FR 32600; June 15, 1998). The double containment requirement is in 
addition to the existing subparts E and F requirements imposed on Type 
B packagings (e.g., the normal conditions of transport and hypothetical 
accident conditions of Secs. 71.71 and 71.73, respectively, and the 
fissile package requirements of Secs. 71.55 and 71.59). Part 71 does 
not impose a double containment requirement for any radionuclide other 
than plutonium. Additionally, IAEA standard TS-R-1 does not provide for 
a double containment requirement (in lieu of the single containment 
Type B package standards) for any radionuclide.
    The AEC issued this regulation at a time when AEC staff anticipated 
widespread reprocessing of commercial spent fuel, and existing 
shipments of plutonium were made in the form of liquid plutonium 
nitrate. Because of physical changes to the plutonium that was expected 
to be reprocessed (i.e., higher levels of burnup in commercial reactors 
for spent fuel, which would then be reprocessed), and regulatory 
concerns with the possibility of package leakage, the AEC issued a 
regulation that imposed the double containment requirement when the 
package contained more than 0.74 TBq (20 Ci) of plutonium. This double 
containment was in addition to the existing Type B package standards on 
packages intended for the shipment of greater than an A1 or A2 quantity 
of plutonium.
    NRC staff has reviewed the available regulatory history for 
Sec. 71.63, and has provided a recapitulation of the supporting 
information which led to the issuance of this regulation. NRC staff has 
extracted the following information from several SECY papers the AEC 
staff submitted to the Commission on this regulation. NRC staff 
believes this information is relevant and will provide stakeholders 
with perspective in understanding the bases for this regulation, and 
thereby assist stakeholders in evaluating the staff's proposed changes 
to this regulation.
    In SECY-R-702,\11\ the AEC staff identified two considerations that 
were the genesis of the rulemaking that led to Sec. 71.63. AEC staff 
stated:
---------------------------------------------------------------------------

    \11\ SECY-R-702, ``Consideration of Form for Shipping 
Plutonium,'' June 1, 1973.

    First, increasingly larger quantities of plutonium will be 
recovered from power reactor spent fuel. Second, the specific 
activity of the plutonium will increase with higher reactor fuel 
burnup resulting in greater pressure generation potential from 
plutonium nitrate solutions in shipping containers, greater heat 
generation, and higher gamma and neutron radiation levels. These 
changes will make the present nitrate packages obsolete. Thus, from 
both safety and economic considerations, the transportation of 
plutonium as [liquid] nitrate will soon require substantial redesign 
of packages to handle larger quantities as well as to deal with the 
higher levels of gas evolution (pressurization), heat generation, 
and gamma and neutron radiation.
    There is little doubt that larger plutonium nitrate packages 
could be designed to meet regulatory standards. The increased 
potential for human error and the consequences of such error in the 
shipment of plutonium nitrate are not so easily controlled by 
regulation. Even though such packages may be adequately designed, 
their loading and closure requires high operation performance by 
personnel on a continuing basis. As the number of packages to be 
shipped increases, the probability of leakage through improperly 
assembled and closed packages also increases.* * * More refined or 
stringent regulatory requirements, such as double containment, would 
not sufficiently lessen this concern because of the necessary 
dependence on people to affect engineered safeguards.

    In SECY-R-74-5,\12\ AEC staff summarized the factors relevant to 
consideration of a proposed rule following a June 14, 1973, meeting to 
discuss SECY-R-702, between the Regulatory and General Manager's staffs 
(i.e., the rulemaking and operational sides of the AEC). The AEC 
stated:
---------------------------------------------------------------------------

    \12\ SECY-R-74-5, ``Consideration of Form for Shipping 
Plutonium,'' dated July 6, 1973.
---------------------------------------------------------------------------

    As a result of this meeting [on June 14, 1973], the [Regulatory and 
General Manager's] staffs have agreed that the basic factors pertinent 
to the consideration of form for shipment of plutonium are:

    1. The experience with shipping plutonium as an aqueous nitrate 
solution in packages meeting current regulatory criteria has been 
satisfactory to date.
    2. The changing characteristic of plutonium recovered from power 
reactors will make the existing packaging obsolete for plutonium 
nitrate solutions and possibly for solid form. Economic factors will 
probably dictate considerably larger shipments (and larger packages) 
than currently used.
    3. It is expected that packages can be designed to meet 
regulatory standards for either aqueous solutions or solid plutonium 
compounds. Just as in any situation involving the packaging of 
radioactive materials, a high level of human performance is 
necessary to assure against leakage caused by human error in 
packaging. As the number of plutonium shipments increases, as it 
will, and packages become larger and more complex in design, the 
probability of such human error increases.
    4. The probability of human error with the packaging for liquid, 
anticipated to be more complex in design, is probably greater than 
with the packaging for solid. Furthermore, should a human error 
occur in package preparation or closure, the probability of liquid 
escaping from the improperly prepared package is greater than for 
most solids and particularly for solid plutonium materials expected 
to be shipped.
    5. Staff studies reported in SECY-R-62 and SECY-R- 509\13\ 
conclude that the consequences of release of solid or aqueous 
solutions do not differ appreciably. Therefore, this paper (SECY-R-
702) does not deal with the consequences of releases.
---------------------------------------------------------------------------

    \13\ SECY-R-62, ``Shipment of Plutonium,'' and SECY-R-509, 
``Plutonium Handling and Storage,'' dated October 16, 1970. These 
papers concluded that there is no scientific or technical reason to 
prohibit shipment of plutonium nitrate and recommended that 
Commission (AEC) efforts be directed toward providing improved 
safety criteria for shipping containers.
---------------------------------------------------------------------------

    6. It is therefore concluded that safety would be enhanced if 
plutonium were shipped as a solid rather than in solution.

    The arguments for requiring a solid form of plutonium for shipment 
are largely subjective, in that there is no hard evidence on which to 
base statistical probabilities or to assess quantitatively the 
incremental increase in safety which is expected. The

[[Page 21422]]

discussion in the regulatory paper, SECY-R-702, is not intended to be a 
technical argument which incontrovertibly leads to a conclusion. It is, 
rather, a presentation of the rationale which has led the Regulatory 
staff to its conclusion that a possible problem may develop and that 
the proposed action is a step towards increased assurance against the 
problem developing. In SECY-R-74-172,\14\ AEC staff submitted a final 
rule to the Commission for approval.
---------------------------------------------------------------------------

    \14\ SECY-R-74-172, ``Consideration of Form for Shipping 
Plutonium,'' April 18, 1974.
---------------------------------------------------------------------------

    The proposed rule had contained a requirement that the plutonium be 
contained in a special form capsule. However, in response to comments 
from the AEC General Manager, the final rule changed this requirement 
to a separate inner container (i.e., the double containment 
requirement). The AEC staff indicated in a response to a public comment 
in Enclosure B (to SECY-R-74-172) that ``[t]he need for the inner 
containment is based on the desire to provide a substitute for not 
requiring the plutonium to be in a `nonrespirable' form.''
    The NRC staff believes the regulatory history of Sec. 71.63 
indicates that the AEC's decision to require a separate inner container 
for shipments of plutonium in excess of 0.74 TBq (20 Ci) was based on 
policy and regulatory concerns (i.e., ``that a possible problem may 
develop and that the proposed action [in SECY-R-702] is a step towards 
increased assurance against the problem developing''). Because of the 
expectation of a significant increase in the number of liquid plutonium 
nitrate shipments, the AEC used a defense-in-depth philosophy (i.e., 
the double containment and solid form requirements), to ensure that 
respirable plutonium would not be released to the environment during a 
transportation accident. However, the regulatory history does indicate 
that the AEC's concerns did not involve the adequacy of existing liquid 
plutonium nitrate packages. Rather, the AEC's regulatory concern was on 
the increased possibility of human error combined with an expected 
increase in the number of shipments would yield an increased 
probability of leakage during shipment. The AEC's policy concern was 
based on an economic decision on whether the AEC should require the 
reprocessing industry to build new, larger liquid plutonium-nitrate 
shipping containers, capable of handling higher burnup reactor spent 
fuel, or to build new, dry, powdered plutonium-dioxide shipping 
containers. The regulatory history indicates that the AEC staff judged 
that new, larger, higher burnup-capacity liquid plutonium-nitrate 
packages could be designed, approved, built, and safely used. However, 
one of the AEC's principal underlying assumptions for this rule was 
obviated in 1979 when the Carter administration decided that 
reprocessing of civilian spent fuel and reuse of plutonium was not 
desirable. Consequently, the expected plutonium reprocessing economy 
and widespread shipments of liquid plutonium nitrate within the U.S. 
never materialized.
    On June 15, 1998, in response to a petition for rulemaking 
submitted by DOE (PRM-71-11), the Commission issued a final rule 
revising Sec. 71.63(b) to add vitrified high-level waste (HLW) 
contained in a sealed canister to the list of forms of plutonium exempt 
from the double containment requirement (June 15, 1998; 63 FR 32600). 
In its original response to PRM-71-11, NRC proposed in SECY-96-215\15\ 
to make a ``determination'' under Sec. 71.63(b)(3) that vitrified HLW 
contained in a sealed canister did not require double containment. 
However, the Commission in an SRM on SECY-96-215, dated October 31, 
1996, disapproved the staff's approach and directed that resolution of 
this petition be addressed through rulemaking (the June 15, 1998, final 
rule was the culmination of this effort). In addition to disapproving 
the use of a ``determination'' process, the Commission also directed 
the staff to ``* * * also address whether the technical basis for 10 
CFR 71.63 remains valid, or whether a revision or elimination of 
portions of 10 CFR 71.63 is needed to provide flexibility for current 
and future technologies.'' In SECY-97-218,\16\ NRC responded to the 
SRM's direction and stated ``[t]he technical basis remains valid and 
the provisions provide adequate flexibility for current and future 
technologies.''
---------------------------------------------------------------------------

    \15\ SECY-96-215, ``Requirements for Shipping Packages Used to 
Transport Vitrified Waste Containing Plutonium,'' dated October 8, 
1996.
    \16\ SECY-97-218, ``Special Provisions for Transport of Large 
Quantities of Plutonium (Response to Staff Requirements Memorandum--
SECY-96-215),'' dated September 29, 1997.
---------------------------------------------------------------------------

    Petition: The NRC received a petition for rulemaking from 
International Energy Consultants, Inc. (IEC), dated September 25, 1997. 
The petition was docketed as PRM-71-12 and was published for public 
comment (63 FR 8362; February 19, 1998). Based on a request from 
General Atomic, the comment period was extended to July 31, 1998 (see 
63 FR 34335; June 24, 1998). Nine public comments were received on the 
petition. Four commenters supported the petition, and five commenters 
opposed the petition.
    The petitioner requested that Sec. 71.63(b) be removed. The 
petitioner argued that the double containment provisions of 
Sec. 71.63(b) cannot be supported technically or logically. The 
petitioner stated that based on the ``Q-system for the Calculation of 
A1 and A2 Values,'' an A2 quantity of 
any radionuclide has the same potential for damaging the environment 
and the human species as an A2 quantity of any other 
radionuclide.
    NRC believes that the Q-values are based upon radiological exposure 
hazard models which calculate the allowable quantity limit (the 
A1 or A2 value) necessary to produce a known 
exposure (i.e., one A2 of plutonium-239 or one A2 
of cobalt-60 will both yield the same radiation dose under the Q-system 
models, even though the A2 values for these nuclides are 
different [e.g., one A2 of plutonium-239 = 2 x 
104 TBq of plutonium and one A2 of cobalt-60 = 1 
TBq of cobalt]). The Q-system models take into account the exposure 
pathways of the various radionuclides, typical chemical forms of the 
radionuclide, methods for uptake into the body, methods for removal 
from the body, the type of radiation the radionuclide emits, and the 
bodily organs the radionuclide preferentially affects. The specific 
A1 and A2 values for each nuclide are developed 
using radiation dosimetry approaches recommended by the World Health 
Organization and the International Commission on Radiological 
Protection (ICRP). The models are periodically reviewed by 
international health physics experts (including representatives from 
the United States), and the A1 and A2 values are 
updated during the IAEA revision process, based upon the best available 
data. (Note that changes to the A1 and A2 values 
as a result of changes to the models in TS-R-1 are also discussed in 
Issue 3.) These values are then issued by the IAEA in safety standards 
such as TS-R-1. When the IAEA has revised the A1 and 
A2 values in previous revisions of its transport 
regulations, these revised values have been adopted by the NRC and DOT 
into the transportation regulations in 10 CFR Part 71 and 49 CFR Part 
173, respectively.
    NRC's review of the current A1 and A2 values 
in Appendix A to Part 71, Table A-1, reveals that 5 radionuclides have 
an A2 value lower than plutonium (i.e., plutonium-239), and 
11 radionuclides have an A2 value that is equal to 
plutonium-239. Because the models used to determine the A1 
and A2 values all result in the same radiation exposure

[[Page 21423]]

(i.e., hazard), a smaller A1 and A2 value for one 
radionuclide would indicate a greater potential hazard to humans than a 
radionuclide with larger A1 and A2 value. Thus, 
the overall Table A-1 can also be viewed as a relative hazard ranking 
(for transportation purposes) of the listed radionuclides. In that 
light, requiring double containment for plutonium alone is not 
consistent with the relative hazard rankings in Table A-1.
    The petitioner also argued that the Type B package requirements 
should be applied consistently for any radionuclide, whenever a 
package's contents exceed an A2 limit. However, Part 71 is 
not consistent by imposing the double containment requirement for 
plutonium. The petitioner believes that if Type B package standards are 
sufficient for a quantity of a particular radionuclide which exceeds 
the A2 limit, then Type B package standards should also be 
sufficient for any other radionuclide which also exceeds the 
A2 limit. The petitioner stated that:

    While, for the most part, Part 71 regulations embrace this 
simple logical congruence, the congruence fails under 10 CFR 
71.63(b) wherein packages containing plutonium must include a 
separate inner container for quantities of plutonium having a 
radioactivity exceeding 20 curies [0.74 TBq] (with certain 
exceptions).
    The petitioner further stated that:

    If the NRC allows this failure of congruence to persist, the 
regulations will be vulnerable to the following challenges: (1) The 
logical foundation of the adequacy of A2 values as a 
proper measure of the potential for damaging the environment and the 
human species, as set forth under the Q-System, is compromised; (2) 
the absence of a limit for every other radionuclide which, if 
exceeded, would require a separate inner container, is an inherently 
inconsistent safety practice; and (3) the performance requirements 
for Type B packages, as called for by 10 CFR Part 71, establish 
containment conditions under different levels of package trauma. The 
satisfaction of these Type B package standards should be a matter of 
proper design work by the package designer and proper evaluation of 
the design through regulatory review. The imposition of any specific 
package design feature such as that contained in 10 CFR 71.63(b) is 
gratuitous. The regulations are not formulated as package design 
specifications, nor should they be.

    NRC agrees that the Part 71 regulations are not formulated as 
package design specifications; rather, the Part 71 regulations 
establish performance standards for a package's design. The NRC reviews 
the application to evaluate whether the package's design meets the 
performance requirements of Part 71. Consequently, the NRC can then 
conclude that the design of the package provides reasonable assurance 
that public health and safety and the environment are adequately 
protected.
    The petitioner also believes that the continuing presence of 
Sec. 71.63(b) engenders excessively high costs in the transport of some 
radioactive materials without a clearly measurable net safety benefit. 
The petitioner stated that this is so, in part, because the ultimate 
release limits allowed under Part 71 package performance requirements 
are identical with or without a ``separate inner container,'' and 
because the presence of a ``separate inner container'' promotes 
additional exposures to radiation through the additional handling 
required for the ``separate inner container.'' Consequently, the 
petitioner asserted that the presence or absence of a separate inner 
container barrier does not affect the standard to which the outer 
container barrier must perform in protecting public health and safety 
and the environment. Therefore, the petitioner concluded that given 
that the outer containment barrier provides an acceptable level of 
safety, the separate inner container is superfluous and results in 
unnecessary cost and radiation exposure. According to the petitioner, 
these unnecessary costs involve both the design, review, and 
fabrication of a package, as well as the costs of transporting the 
package. And the unnecessary radiation exposure involves workers having 
to handle (i.e., seal, inspect, or move) the ``separate inner 
container.''
    As an alternative to the primary petition, the petitioner believes 
that an option to eliminate both Sec. 71.63(a) and (b) should also be 
considered. Section 71.63(a) requires that plutonium in quantities 
greater than 0.74 TBq (20 Ci) be shipped in solid form. This option 
would have the effect of removing Sec. 71.63 entirely. The petitioner 
believes that the arguments set forth to support the elimination of 
Sec. 71.63(b) also support the elimination of Sec. 71.63(a). The 
petitioner did not provide a separate regulatory or cost analysis 
supporting the request to remove Sec. 71.63(a).
    Comments on the Petition: The four commenters supporting the 
petition essentially stated that the IAEA's Q-system accurately 
reflects the dangers of radionuclides, including plutonium, and that 
elimination of Sec. 71.63(a) and (b) would make the regulations more 
performance based, reduce costs and personnel exposures, and be 
consistent with the IAEA standards.
    The five commenters opposing the petition essentially stated that: 
(1) Plutonium is very dangerous, especially in liquid form, and 
therefore additional regulatory requirements are warranted; (2) 
Existing regulations are not overly burdensome, especially in light of 
the total expected transportation cost; (3) TRUPACT-II packages meet 
current Sec. 71.63(b) requirements (TRUPACT-II is a package developed 
by DOE to transport transuranic wastes (including plutonium) to the 
Waste Isolation Pilot Plant (WIPP) and has been issued a Part 71 CoC, 
No. 9218); (4) A commenter (the Western Governors' Association) has 
worked for over 10 years to ensure a safe transportation system for 
WIPP, including educating the public about the TRUPACT-II package; (5) 
Any change now would erode public confidence and be detrimental to the 
entire transportation system for WIPP shipments; and (6) Additional 
personnel exposure due to double containment is insignificant.
    Discussion: The NRC has received 48 public comments on this issue 
in response to the Issue Paper, public meetings, and the workshop. 
Industry representatives and some members of the public support the 
petition. Public interest organizations, Agreement States, State 
representatives, the Western Governors' Association, and other members 
of the public oppose the petition. Several commenters believe that 
Congress, in approving the Waste Isolation Pilot Plant Land Withdrawal 
Act (the Act), Pub. L. 102-579 (106 Stat. 4777), Section 16(a), which 
mandates that the NRC certify the design of packages used to transport 
transuranic waste to WIPP, expected those packages to have a double 
containment. The NRC researched this issue, and Section 16(a) of the 
Act does not contain any explicit provisions mandating the use of a 
double containment in packages transporting transuranic waste to or 
from WIPP. Section 16(a) of the Act states, in part, ``[n]o transuranic 
waste may be transported by or for the Secretary [of the DOE] to or 
from WIPP, except in packages the design of which has been certified by 
the Nuclear Regulatory Commission* * *'' Furthermore, the NRC has 
reviewed the legislative history\17\ associated with the Act and has 
not identified any discussions on the use of double containment for the 
shipment of

[[Page 21424]]

transuranic waste. The legislative history does mention that the design 
of these packages will be certified by the NRC; however, this language 
is identical to that contained in the Act itself. Therefore, the NRC 
believes the absence of specific language in Section16(a) of the Act 
requiring double containment should be interpreted as requiring the NRC 
to apply its independent technical judgment in establishing standards 
for package designs and in evaluating applications for certification of 
package designs, to ensure that such packages would provide reasonable 
assurance that public health and safety and the environment would be 
adequately protected. In carrying out its mission, the courts have 
found that the NRC has broad latitude in establishing, maintaining, and 
revising technical performance criteria necessary to provide reasonable 
assurance that public health and safety and the environment are 
adequately protected. An example of these technical performance 
criteria is the Type B package design standards. Accordingly, the NRC 
believes that the proposed revision of a technical package standard 
(i.e., removal of the double containment requirement for plutonium from 
the Type B package standards) is not restricted by the mandate of 
Section 16(a) of the Act for the NRC to certify the design of packages 
intended to transport transuranic material to and from WIPP.
---------------------------------------------------------------------------

    \17\ See Congressional Record Vol. 137, November 5, 1991, pages 
S15984--15997 (Senate approval of S. 1671); Cong. Rec. Vol. 138, 
July 21, 1992, pages H6301--6333 (House approval of H.R. 2637); 
Cong. Rec. Vol. 138, October 5, 1992, pages H11868--11870 (House 
approval of Conference Report on S. 1671); Cong. Rec. Vol. 138, 
October 8, 1992 (Senate approval of Conference Report on S. 1671); 
and Cong. Rec. Vol. 138, October 5, 1992, pages H12221--12226 
(Conference Report on S. 1671--(H.) Rpt. 102-1037).
---------------------------------------------------------------------------

    Other commenters stated that stakeholders' expectations were that 
packages intended to transport transuranic material to and from WIPP 
would include a double containment provision. Consequently, the 
commenters believed that removal of the double containment requirement 
would decrease public confidence in the NRC's accomplishment of its 
mission in the approval of the design of packages for the 
transportation of transuranic waste to and from WIPP. The commenters 
believed the public would view elimination of the double containment 
requirement as a relaxation in safety. The presence of a separate inner 
container provides defense-in-depth through an additional barrier to 
the release of plutonium during a transportation accident. In addition, 
the commenters believed that plutonium is so inherently deadly, that 
defense-in-depth is appropriate. The NRC agrees that a double 
containment does provide an additional barrier. However, the NRC 
believes that, for the reasons discussed below, double containment is 
unnecessary to protect public health and safety. The NRC and AEC have 
not required an additional containment barrier for Type B packages 
transporting any radionuclides other than plutonium and, before 1974, 
the AEC did not require double containment for plutonium.
    In response to some of the comments opposed to the petition, the 
NRC believes that removal of Sec. 71.63(b) would not invalidate the 
design of existing packages intended for the shipment of plutonium. 
These packages could continue to be used with a separate inner 
container. The NRC agrees with the commenters that a quantitative cost 
analysis was not provided by the petitioner.
    The NRC has issued Part 71 CoC No. 9218 to DOE for the TRUPACT-II 
package (Docket No. 71-9218), for the transportation of transuranic 
waste (including plutonium) to and from the WIPP. The TRUPACT-II 
package complies with the current Sec. 71.63(b) requirements and has a 
separate inner container. The TRUPACT-II SAR indicates that the weight 
of the inner container and its lid is approximately 2,620 lbs. 
Hypothetically, elimination of the separate inner container would 
increase the available payload for the TRUPACT-II package from the 
current 7,265 to 9,885 lbs. Thus, removal of the double containment 
requirement would potentially increase the TRUPACT-II's available 
payload by 36 percent. Further, the removal of the inner container from 
the TRUPACT-II would also potentially increase the available volume. 
The NRC believes that the proposed rule would not invalidate the 
existing TRUPACT-II design, and thus, DOE could continue to use the 
TRUPACT-II to ship transuranic waste to and from WIPP, or DOE could 
consider an alternate Type B package.
    Additionally, based on comments received in the public meetings, 
the NRC believes that a misperception exists with respect to TRUPACT-II 
shipments; removal of the Sec. 71.63(b) double containment requirement 
would not result in loose plutonium waste being placed inside a 
TRUPACT-II package. Based upon information contained in the SAR, 
plutonium wastes (i.e., used gloves, anti-Cs, rags, etc.) are placed in 
plastic bags, and these bags are sealed inside lined 55-gallon steel 
drums. Plutonium residues are placed inside cans which are then sealed 
inside a pipe overpack (a 6-inch or 12-inch stainless steel cylinder 
with a bolted lid), and the pipe overpack is then sealed inside a lined 
55-gallon steel drum. The 55-gallon drums are then sealed inside the 
TRUPACT-II inner containment vessel, and finally the inner containment 
vessel is sealed inside the TRUPACT-II package. Consequently, the 
TRUPACT-II shipping practices employ multiple barriers, and removal of 
the inner containment vessel would not be expected to produce a 
significant incremental increase in the possibility of leakage during 
normal transportation. The NRC notes that some NRC regulations have 
established additional requirements for plutonium (e.g., the special 
nuclear material license application provisions of Sec. 70.22(f)).
    The NRC believes that the Type B packaging standards, in and of 
themselves, provide reasonable assurance that public health and safety 
and the environment would be adequately protected during the 
transportation of radioactive material. This belief is supported by an 
excellent safety record in which no fatalities or injuries have been 
attributed to material transported in a Type B package. Type B 
packaging standards have been in existence for approximately 40 years 
and have been incorporated into the Part 71 regulations by both the NRC 
and its predecessor, the AEC. The NRC's Type B package standards are 
based on IAEA's Type B package standards. Moreover, IAEA's Type B 
package standards have never required a separate inner container for 
packages intended to transport plutonium, nor for any other 
radionuclide. The NRC believes that while U.S. shipments of plutonium 
subject to Sec. 71.63(b) have consisted primarily of solid plutonium 
contaminated wastes, other European countries have reprocessed 
plutonium in their reactor fuel cycles and have transported liquid 
plutonium nitrate. The NRC is not aware of any accidents involving a 
Type B liquid plutonium nitrate package which has led to the 
significant failure of the package and release of the contents.
    Therefore, the NRC believes that imposition of an additional 
packaging requirement (in the form of a separate inner container) is 
fundamentally inconsistent with the position that Type B packaging 
standards, in and of themselves, provide reasonable assurance that 
public health and safety and the environment would be adequately 
protected during the transportation of (any type of) radioactive 
material. Thus, the NRC believes that Sec. 71.63(b) is not consistent 
with the Type B packaging standards contained in part 71.
    The NRC also believes that the regulatory history of Sec. 71.63 
demonstrates that the AEC's decision was based on policy and regulatory 
concerns. However, the NRC also agrees that the use of a double 
containment

[[Page 21425]]

does provide defense-in-depth and does decrease the absolute risk of 
the release of respirable plutonium to the environment during a 
transportation accident. Consequently, while the defense-in-depth 
afforded by a double containment does reduce risk, the NRC believes the 
question which should be focused on is whether the double containment 
requirement is risk-informed. The NRC is unaware of any risk studies 
that would provide either a qualitative or quantitative indication of 
the risk reduction associated with the use of double containment in 
transportation of plutonium. Rather, the NRC would look to the 
demonstrated performance record of existing Type B package standards to 
conclude that double containment is not necessary.
    In summary, the AEC indicated (in SECY-R-702 and SECY-R-74-5), that 
liquid plutonium nitrate packages were safe, and new, larger packages 
to handle higher burnup reactor spent fuel could also be designed. NRC 
believes that the AEC's assumption for initiating this requirement was 
that large scale reprocessing of civilian reactor spent fuel and reuse 
of plutonium would occur. Former President Carter's administration's 
decision to forgo the reprocessing of civilian reactor spent fuel and 
reuse of plutonium obviated the AEC's assumption. Consequently, the 
AEC's supposition that a human error occurring while sealing a package 
of liquid plutonium nitrate was more likely to occur with the expected 
increase in shipments of plutonium nitrate was also obviated by the 
Government's decision to forgo the reprocessing of civilian reactor 
spent fuel. In SECY-97-218, NRC staff indicated that the separate inner 
container provided an additional barrier to the release of plutonium in 
an accident. NRC continues to believe that a separate inner container 
provides an additional barrier to the release of plutonium in an 
accident, just as a package with triple containment would provide an 
even greater barrier to the release of plutonium in an accident. 
However, this type of approach is not risk informed nor performance 
based. Consequently, based upon review of the petition, comments on the 
petition, and research into the regulatory history of the double 
containment requirement, the NRC agrees that a separate inner container 
is not necessary for Type B packages containing solid plutonium. NRC 
believes that the worldwide performance record over 40 years of Type B 
packages demonstrates that a single containment barrier is adequate. 
Therefore, the NRC agrees with the petitioner and believes that 
Sec. 71.63(b) is not technically necessary to provide a reasonable 
assurance that public health and safety and the environment will be 
adequately protected during the transportation of plutonium.
    While the NRC believes a case can be made for elimination of the 
separate inner container requirement in Sec. 71.63(b), elimination of 
the solid form requirement in Sec. 71.63(a) is not as clear. While the 
same arguments can be made on the obviation of the AEC's basis for 
originally issuing Sec. 71.63(a) (i.e., the elimination of reprocessing 
of plutonium), the same regulatory inconsistency between Type B package 
standards and the inner containment requirement does not exist for the 
liquid versus solid form argument. The NRC considers the contents of a 
package when it is evaluating the adequacy of a packaging's design. The 
approved content limits and the approved packaging design together 
define the CoC for a package. However, other than criticality controls 
and the liquid form requirement of Sec. 71.63(a), Subparts E and F do 
not contain any restrictions on the contents of a package. Thus, while 
the inner containment requirement in Sec. 71.63(b) can be seen as 
conflicting with the Type B package standard because the inner 
containment affects the packaging's design, the solid form requirement 
of Sec. 71.63(a) does not conflict with the packaging requirements of 
the Type B package standard because the solid form requirement affects 
only the contents of the package, not the packaging itself.
    The NRC expects that cost and dose savings would accrue from the 
removal of Sec. 71.63(b). However, because no shipments of liquid 
plutonium nitrate are contemplated in the U.S., NRC does not expect 
cost or dose savings to accrue from the removal of Sec. 71.63(a). 
Further, the AEC's original bases have been obviated by former 
President Carter's administration's decision to not pursue a commercial 
fuel cycle involving the reprocessing of plutonium.
    After weighing this information, the NRC continues to believe that 
the Type B package standards, when evaluated against 40 years of use 
worldwide, and millions of safe shipments of Type B packages, together 
provide reasonable assurance that public health and safety and the 
environment would be adequately protected during the transportation of 
radioactive material. The NRC believes that, in this case, the 
reasonable assurance standard, provided by the Type B package 
requirements, provides an adequate basis for the public's confidence in 
the NRC's actions.
    NRC Proposed Position: The NRC would adopt, in part, the 
recommended action of PRM-71-12. Specifically, the NRC would remove the 
double containment requirement of Sec. 71.63(b). However, the NRC would 
retain the package contents requirement in Sec. 71.63(a). Shipments 
whose contents contain greater than 0.74 TBq (20 Ci) of plutonium must 
be made with the contents in solid form.
    Affected Sections. Sec. 71.63.
Issue 18. Contamination Limits as Applied to Spent Fuel and High Level 
Waste (HLW) Packages
    Background. In the period of December 1997 through April 1998, the 
French Nuclear Installations Safety Directorate inspected a French 
nuclear power plant and railway terminal used by the La Hague 
reprocessing plant. The inspectors noticed that, since the beginning of 
the 1990's, a high percentage of spent fuel packages and/or railcars 
had a level of removable surface contamination that exceeded IAEA 
regulatory limits by as much as a factor of 1000. Subsequent 
investigations found that the contamination incidents involved 
shipments from other European countries, and the French transport 
authorities notified their counterparts of their findings. 
Subsequently, French, German, Swiss, Belgian, and Dutch spent fuel 
shipments were temporarily suspended.
    After estimating the occupational and public doses from the 
contamination incidents, the European transport authorities concluded 
that these incidents did not have any radiological consequence. The 
contamination was believed to be caused by contact of the spent fuel 
package surface with contaminated water from the spent fuel storage 
pool during package handling operations. The authorities concluded that 
there were deficiencies in the contamination measurement procedures and 
the distribution of that information.
    Media reports on these incidents focused attention on IAEA's 
regulations for removable contamination on package surfaces. TS-R-1 
contains contamination limits for all packages of 4.0 Bq/cm2 
for beta and gamma and low toxicity alpha emitting radionuclides, and 
0.4 Bq/cm2 for all other alpha emitting radionuclides. 
Although TS-R-1 uses the term limit, IAEA considers these ``limits'' to 
be guidance values, or derived values, above which appropriate action 
should be considered. In cases of contamination above the limit, that 
action is to decontaminate to below the limits.

[[Page 21426]]

    The current TS-R-1 limits for removable package surface 
contamination were derived from a radiological model developed for the 
1961 Edition of the IAEA regulations. The exposure pathways considered 
in the model included external irradiation of the skin, and ingestion 
and inhalation from resuspension of the contamination in air. The model 
uses values for the degree to which surface contamination is 
resuspended in air, making it available for inhalation, and for the 
number of hours of exposure to the resuspended contamination. The 
values were chosen to represent occupational conditions at shipper and 
carrier facilities, in which workers manually handled many packages 
throughout the year. These exposure conditions are much greater than 
the public would experience from brief exposure to packages in 
transport. The values also exceed real occupational resuspension rates 
and exposure times and were believed to result in worker doses that 
would be well within the annual occupational dose limit. Exposure at 
the contamination limit does not pose a significant health hazard to 
workers. Therefore, members of the public, few of whom would ever be 
expected to encounter contaminated packages in transit, and then only 
briefly, are also protected against contamination hazards by the limit.
    TS-R-1 further provides that in transport, ``* * * the magnitude of 
individual doses, the number of persons exposed, and the likelihood of 
incurring exposure shall be kept as low as reasonable, economic and 
social factors being taken into account * * *'' The IAEA contamination 
regulations have been applied to radioactive material packages in 
international commerce for almost 40 years, and practical experience 
demonstrates that the regulations can be applied successfully. With 
respect to contamination limits, TS-R-1 contains no changes from 
previous versions of IAEA's regulations.
    Part 71 does not contain contamination limits, but Sec. 71.87(i) 
requires that licensees determine that the level of removable 
contamination on the external surface of each package offered for 
transport is as low as is reasonably achievable, and within the limits 
specified in DOT regulations in 49 CFR 173.443. The DOT contamination 
limits differ from TS-R-1 in that the contamination limits apply to the 
wipe material used to survey the surface of the package, not the 
surface itself. Also, the contamination limits are only 10 percent of 
the TS-R-1 values (e.g., wipe limit of 0.4 Bq/cm2 (2200 dpm/
100 cm2 ) for beta and gamma and low toxicity alpha emitting 
radionuclides), because the DOT limits are based on the assumption that 
the wipe removes 10 percent of the surface contamination. In this 
regard, the DOT and TS-R-1 limits are equivalent.
    The DOT contamination regulations contain an additional provision 
for which there is no counterpart in TS-R-1. Section 173.443(b) 
provides that, for packages transported as exclusive use (see 49 CFR 
173.403 for exclusive use definition) shipments by rail or public 
highway only, the removable contamination on any package at any time 
during transport may not exceed 10 times the contamination limits 
(e.g., wipe contamination of 4 Bq/cm2 (22,000 dpm/100 
cm2) for beta and gamma and low toxicity alpha emitting 
radionuclides). In practice, this means that packages transported as 
exclusive use shipments (this includes spent fuel packages) that meet 
the contamination limits at shipment departure may have 10 times that 
contamination upon arrival at the destination. This provision is 
intended to address a phenomenon known as ``cask-weeping,'' in which 
surface contamination that is nonremovable at the beginning of a 
shipment becomes removable during the course of the shipment. 
Nonremovable contamination is not measurable using wipe surveys and is 
not subject to the removable contamination limits. At the destination 
facility, a package exhibiting cask-weeping can exceed the 
contamination limits by a considerable margin, even though the package 
met the limits at the originating facility, and was not subjected to 
any further contamination sources during shipment. Environmental 
conditions are believed to affect the cask-weeping phenomenon.
    Spent fuel packages and shipments differ from those considered in 
the 1961 model used to develop package surface contamination limits. 
Workers are exposed to only a few spent fuel packages per year at most, 
so their exposure time to package contamination is less than that 
modeled. Unlike the packages in the model, however, spent fuel package 
surface areas and radiation levels are significant. Exposure to the 
package radiation level while performing either contamination survey or 
decontamination activities contributes to worker dose, and this impact 
was not considered in the model.
    The IAEA has plans to establish a Coordinated Research Project 
(CRP) to review contamination models, approaches to reduce package 
contamination, strategies to address cask-weeping, and possible 
recommendations for revisions to the contamination standard that 
consider risks, costs, and practical experience. IAEA establishes CRPs 
to facilitate investigation of radioactive material transportation 
issues by key IAEA Member States. IAEA will then consider a CRP report 
and any further actions or remedies that may be warranted at periodic 
meetings (at TRANSSC). NRC informed IAEA that NRC supports the IAEA 
initiative to establish the CRP and that NRC would participate in the 
IAEA review of surface contamination standards.
    Discussion. During the three public meetings, NRC has received 
verbal public comments on the contamination issue. One commenter agreed 
that external contamination on packages of radioactive material in 
transport is a significant problem and is the source of actual or 
perceived hazard that can cause damage to the nuclear industry. The 
commenter would prefer not to change contamination limits (i.e., 
continuing to use TS-R-1 limits) unless there is a sound technical 
basis for doing so.
    NRC was requested to clarify its discussion of the 4 Bq/
cm2. The commenter stated that the current limit for 
removable contamination levels in 49 CFR 173.443 is 0.4 Bq/
cm2 before shipment, unless an assessment method with higher 
efficiency is used, in which case the limit may be as high as 10 times 
0.4 Bq/cm2 (i.e., 4 Bq/cm2) (22,000 dpm/100 
cm2 ).
    Four commenters stated they understood that existing surface 
contamination limits (i.e., 4 Bq/cm2) (2200 dpm/100 
cm2) were intended for small and not large packages and that 
using the limit for large packages, while it may reduce public exposure 
rates, would conceivably increase worker exposure rates. Another 
commenter added that worker exposure could actually increase when 
double containment is required, and expressed concern about how this 
issue with contamination limits impacts international shipments. Some 
commenters stated that it was doubtful that worker exposure rates could 
be reduced, even if allowable surface contamination rates were 
significantly increased.
    Several commenters addressed the issue that workers would be 
exposed to radiation while measuring the surface contamination level. 
Three of the commenters acknowledged that this is true regardless of 
the level of the package contamination limit. Two commenters suggested 
that NRC consider other ways to protect workers, including cask design. 
Another commenter stated that if the radiation is

[[Page 21427]]

too great for workers to get close enough to measure it, it is too 
great to transport it.
    Absent public objection to the current standard and an overall 
significantly improved approach, NRC is planning no revisions to Part 
71 regarding surface contamination in this proposed rule. The NRC 
intends to use the information it collects from public comments on this 
issue to continue to support DOT in U.S. participation in the IAEA CRP 
and to work with DOT and other IAEA Member States on this issue. 
Because IAEA has adopted a 2-year revision cycle for TS-R-1, a revision 
based on the CRP's results could be incorporated into TS-R-1 more 
quickly than under the previous 10-year revision cycle.
    NRC Proposed Position. The NRC proposes no changes to Part 71 for 
this issue.
    Affected Sections. None (not adopted).
Issue 19. Modifications of Event Reporting Requirements
    Background. The Commission recently issued a final rule to revise 
the event reporting requirements in 10 CFR Part 50 (see 65 FR 63769; 
October 20, 2000). This final rule revised the verbal and written event 
notification requirements for power reactor licensees in Secs. 50.72 
and 50.73. In SECY-99-181,\18\ NRC staff informed the Commission that 
public comments on the proposed Part 50 rule had suggested that 
conforming changes also be made to the event notification requirements 
in Part 72 (Licensing Requirements for the Independent Storage of Spent 
Fuel) and Part 73 (Physical Protection of Plants and Materials). In 
response, the Commission directed the NRC staff to study whether 
conforming changes should be made to Parts 72 and 73. During this 
study, the NRC also reviewed the Part 71 event reporting requirements 
in Sec. 71.95, and concluded that similar changes could be made to the 
Part 71 event reporting requirements.
---------------------------------------------------------------------------

    \18\ SECY-99-181, ``Proposed Plans and Schedules to Modify 
Reporting Requirements Other than 10 CFR 50.72 and 50.73 for Power 
Reactors and Material Licensees,'' dated July 9, 1999.
---------------------------------------------------------------------------

    Discussion. This issue was not included in the Part 71 Issues Paper 
(65 FR 44360; July 17, 2000). Therefore, there were no public comments 
on this issue.
    The current regulations in Sec. 71.95 require that a licensee 
submit a written report to the NRC within 30 days of three events: (1) 
A significant decrease in the effectiveness of a packaging while it is 
in use to transport radioactive material; (2) details of any defects 
with safety significance found after first use of the cask; and (3) 
failure to comply with conditions of the CoC during use.
    The NRC has identified three principal concerns with the existing 
requirements in Sec. 71.95. First, the existing requirements only apply 
to licensees and not to certificate holders. Second, the existing 
requirements do not contain any direction on the content of these 
written reports. Third, inconsistencies existed in reporting time 
frames as a result of the Commission decision in the October 20, 2000, 
final rule which reduced the reporting burden on reactor licensees in 
the Part 50 final rule by changing the time for submittal of written 
reports from 30 days to 60 days.
    With respect to the first concern, NRC believes that events 
involving a significant reduction in effectiveness of a packaging 
during its use to transport radioactive material may call into question 
the design bases for the packaging. Examples of a significant reduction 
in effectiveness might involve an event that causes a package to exceed 
the 2-mSv per hour (200-mrem per hour) dose limit or exceed the Type B 
package requirements of Sec. 71.51. In these cases, the cause of the 
reduction in effectiveness may be due to a design flaw. Because the 
certificate holder has the most in-depth understanding of the design 
basis for a packaging, the NRC believes that it is appropriate for the 
certificate holder to work with the licensee to jointly determine the 
root cause(s) for an event that resulted in a significant decrease in 
packaging effectiveness. Similarly, identification of safety-
significant defects after first use of a packaging may reveal flaws 
with the packaging's basic design. Therefore, the NRC would revise 
Sec. 71.95 to require that the licensee request certificate holder 
input before submitting a written report for the criteria in new 
paragraphs (a)(1) and (a)(2). The licensee would also be required to 
provide the certificate holder with a copy of the written event report, 
after the report is submitted to the NRC. This would permit the 
certificate holder to monitor and trend package performance information 
arising from package use by multiple licensees. In new paragraph 
(a)(3), the NRC would retain the existing requirement for licensees to 
report instances of failure to follow the conditions of the CoC while a 
packaging was in use.
    With respect to the second concern, NRC believes that direction 
should be provided on the expected contents of these written reports. 
Currently, no direction is provided to licensees on the form or content 
of these written reports. The NRC believes that standards for the 
contents of written reports should be unambiguous. The NRC uses this 
information to determine if inspection and enforcement follow-up is 
required for the event or if a generic safety issue exists. 
Consequently, sufficient information must be provided to the NRC to 
fulfill its responsibilities to protect public health and safety and 
the environment. Therefore, NRC would add new paragraphs (c) and (d) to 
Sec. 71.95 which would provide guidance on the content of these written 
reports. This new requirement is consistent with the written report 
requirements for Parts 50 and 72 licensees (i.e., Secs. 50.73 and 
72.75) and the direction from the Commission in SECY-99-181 to consider 
conforming event notification requirements to the recent changes made 
to Part 50. The NRC would also update the submission location for the 
written reports from the Director, Office of Nuclear Material Safety 
and Safeguards, to the NRC Document Control Desk. This action is 
consistent with previous Commission direction to standardize the 
location for incoming documents and correspondence and would bring Part 
71 into greater conformity with Parts 50 and 72. Additionally, the NRC 
would remove the specific location for submission of written reports 
from Sec. 71.95(c) and require that reports be submitted in accordance 
with Sec. 71.1. This action is also consistent with the approach taken 
in Parts 50 and 72 and would reduce future NRC burden should the 
submission address change. This proposed change to Sec. 71.1 is 
identical to a change made to Sec. 72.4 in a recent Part 72 final rule 
(see 64 FR 33178; June 22, 1999).
    With respect to the third concern, the NRC staff believes that 
lengthening the period for submitting reports from 30 days to 60 days 
would reduce the burden on licensees, while still providing the staff 
with the necessary information to fulfill the NRC's mission. The NRC 
uses written event reports for trending, analysis, and long-term 
follow-up of a licensee's corrective actions. In contrast, immediate 
reporting of events to the NRC provides indication of significant 
events when immediate action to protect public health and safety may be 
required or where the NRC needs timely and accurate information to 
respond (see 48 FR 39039; August 29, 1983, on the basis for Part 50 
event reporting). For transportation events, the NRC receives early 
notification in the NRC's Operations Center either from a

[[Page 21428]]

licensee, when a licensee declares an emergency under its emergency 
plan, for a transportation event, or from DOT's National Response 
Center, when a shipper notifies DOT of an accident involving 
radioactive material. Consequently, extending the submission time for 
written event reports to 60 days would not adversely affect the NRC's 
ability to promptly respond to an event, because these written reports 
are not used as the basis for immediate or short term actions.
    The Commission concluded in the October 20, 2000 (65 FR 63769), 
final rule revising Part 50 event reporting requirements that the 
length of time to submit a written report should be extended to permit 
a thorough evaluation of the event, identification of the root causes, 
and development of corrective actions. The Commission also indicated 
that a licensee's submission of written reports should not be 
unnecessarily delayed to take advantage of the full 60-day period. The 
NRC took this action because some events required a significant amount 
of time to evaluate the event, identify the root causes, and identify 
the corrective actions; and consequently, a supplemental written event 
report was necessary. In addition, a 60-day period is more consistent 
with the NRC's desire that the licensee and the certificate holder both 
be involved in the analysis of an event. The Commission indicated that 
the licensee's burden, in submitting a supplemental written event 
report, would be reduced by providing sufficient time to complete the 
original written event report.
    The NRC staff believes the Commission's rationale for lengthening 
the reporting period from 30 days to 60 days for Part 50 written event 
reports is also valid for Part 71 written event reports.
    The NRC draft RA indicates that adoption of the conforming change 
to Part 71 for event reporting requirements is appropriate from a 
safety, regulatory, and cost perspective. Regulatory efficiency within 
NRC would increase with adoption of this proposed change and would 
result in greater conformity among Parts 50, 71, and 72. Further, NRC 
burden (and thus costs) would be reduced should the submission address 
change in the future. There would be a one-time implementation cost for 
licensees for revising procedures and for training. A key benefit of 
the proposed amendments would be a reduction in the recurring annual 
reporting burden on licensees, as a result of reducing the efforts 
associated with reporting events of little or no risk or safety 
significance. It is anticipated that the NRC's recurring annual review 
efforts for telephone notifications and written reports would not be 
significantly reduced.
    NRC Proposed Position. The NRC proposes a reduction in regulatory 
burden for licensees by lengthening the Sec. 71.95 event reporting 
submission period from 30 to 60 days.
    Affected Sections. Sec. 71.95.

V. Section-by-Section Analysis

    Several sections In Part 71 would be redesignated in this 
rulemaking to improve consistency and ease of use. For some sections, 
only the section number would be changed. However, for other sections, 
revisions would also be made to the regulatory language. The following 
table is provided to aid the public in understanding the proposed 
numerical changes to sections of Part 71.

                           Redesignation Table
------------------------------------------------------------------------
            New section number                 Existing section number
------------------------------------------------------------------------
Sec.  71.8................................  Sec.  71.11
Sec.  71.9................................  New section
Sec.  71.10...............................  New section
Sec.  71.11 (Reserved)....................  NA
Sec.  71.12...............................  Sec.  71.8
Sec.  71.13...............................  Sec.  71.9
Sec.  71.14...............................  Sec.  71.10
Sec.  71.15...............................  Sec.  71. 53
Sec.  71.16 (Reserved)....................  NA
Sec.  71.17...............................  Sec.  71.12
Sec.  71.18...............................  New section
Sec.  71.19...............................  Sec.  71.13
Sec.  71.20...............................  Sec.  71.14
Sec.  71.21...............................  Sec.  71.16
Sec.  71.22...............................  Sec.  71.18
Sec.  71.23...............................  Sec.  71.20
Sec.  71.24 (Reserved)....................  Sec.  71.22 (Section
                                             removed)
Sec.  71.25 (Reserved)....................  Sec.  71.24 (Section
                                             removed)
Sec.  71.53 (Reserved)....................  Sec.  71.53 (Section
                                             redesignated)
------------------------------------------------------------------------

Subpart A--General Provisions

10 CFR 71.0  Purpose and Scope
    Paragraph (d) would be reformatted into four paragraphs to simplify 
this regulation, to better use plain language, and to reflect the 
existence of the new Type B(DP) package approval process in new Subpart 
I. Paragraph (d)(1) would indicate that general licenses for which no 
NRC package approval is required are issued in new Secs. 71.20 through 
71.23. This is changed from the current sentence, because of the 
removal of existing Secs. 71.22 and 71.24 (redesignated Secs. 71.24 and 
71.25). A new sentence would be added referring to the requirement for 
a CoC to be issued for a Type B(DP) package to be used under the new 
general license in new Sec. 71.18. Paragraph (d)(2) would indicate that 
an application for package approval--for package types other than Type 
B(DP)--must be completed in accordance with Subpart D. Paragraph (d)(3) 
would indicate that an application for a Type B(DP) package must be 
completed in accordance with Subpart I. Paragraph (d)(4) would continue 
to require a licensee transporting, or delivering material to a carrier 
for transport, to meet the requirements of the applicable portions of 
Subparts A, G, and H.
    New paragraph (e) would be added to indicate that persons who hold, 
or apply for, a Part 71 CoC for Type AF, Type B, Type BF, Type B(U)F, 
Type B(M)F, and Type B(DP) packages are within the scope of Part 71 
regulations.
    Existing paragraphs (e) and (f) would be redesignated as new 
paragraphs (f) and (g), respectively. The rule text in new paragraph 
(f) would be the same as existing paragraph (e) text. New paragraph (g) 
would be revised to reflect the redesignation of existing Sec. 71.11 as 
new Sec. 71.8.
10 CFR 71.1  Communications and Reports
    In Sec. 71.1, paragraph (a) would be revised to indicate that 
documents submitted to the NRC should be addressed to the attention of 
the ``Document Control Desk,'' not the ``Director of the Office of 
Nuclear Material Safety and Safeguards.'' Provisions would also be 
added to provide requirements when a due date for a document falls on a 
Saturday, Sunday, or Federal holiday. In that case, the document would 
be due the next Federal work day. This change would be identical to a 
change made to Sec. 72.4 in a recent Part 72 final rule (see 64 FR 
33178; June 22, 1999).
10 CFR 71.2  Interpretations
    No changes were made to the text of this section; however, it is 
included in the revision of this subpart for completeness.
10 CFR 71.3  Requirement for License
    No changes were made to the text of this section; however, it is 
included in the revision of this subpart for completeness.
10 CFR 71.4  Definitions
    The existing definitions for A1, Fissile material, Low Specific 
Activity (LSA) material, Package, and Transport index (TI) would be 
revised as conforming changes. New definitions for A2, 
Certificate of Compliance, Criticality Safety Index (CSI), Deuterium, 
DOT, Graphite, Spent fuel, and Structures,

[[Page 21429]]

systems, and components important to safety would be added as 
conforming changes.
    The definition of A1 would be revised to split the 
current combined definition for A1 and A2 into 
two individual definitions. This approach is consistent with standard 
in TS-R-1. Furthermore, no change would be made to the current 
technical content of the definition for A1; however, the text would be 
revised to improve readability.
    A definition for A2 would be added, because the current 
joint definition for A1 and A2 would be split 
into two definitions. [See also definition for A1.]
    A definition for Certificate of Compliance would be added. This 
definition would be similar to the definition for the same term found 
in Sec. 72.3.
    A definition of Criticality Safety Index (CSI) would be added.
    A definition of Deuterium would be added to indicate that, for the 
purposes of new Secs. 71.15 and 71.22, the definition of ``deuterium'' 
found in 10 CFR 110.2 applies.
    A definition of U.S. Department of Transportation (DOT) would be 
added.
    The definition of Fissile material would be revised by removing 
\238\Pu from the list of fissile nuclides; clarifying that ``fissile 
material'' means the fissile nuclides themselves, not materials 
containing fissile nuclides; and redesignating the reference to 
exclusions from fissile material controls from Sec. 71.53 to new 
Sec. 71.15.
    A definition of Graphite would be added to indicate that, for the 
purposes of new Secs. 71.15 and 71.22, the definition of Nuclear grade 
graphite found in Sec. 110.2 applies.
    The definition of Low Specific Activity (LSA) material , for LSA-
III material, would be revised to reflect the existence of Sec. 71.77 
(Sec. 71.77 provides requirements on the qualification of LSA-III 
material).
    The definition of Package would be revised by clarifying in 
paragraph (1) that Fissile material package also means a Type AF, Type 
BF, Type B(U)F, or Type B(M)F package. New paragraph (2) would be added 
defining Type A packages in accordance with DOT regulations contained 
in 49 CFR Part 173. Existing paragraph (2) defining Type B packages 
would be redesignated as paragraph (3). No changes would be made to the 
redesignated text. New paragraph (4) would be added defining a Type 
B(DP) package.
    A definition of Spent nuclear fuel or Spent fuel would be added. 
This definition is the same as that currently found in Sec. 72.3.
    A definition for Structures, systems, and components important to 
safety would be added for Type B(DP) packages. This definition would be 
similar to the definition currently found in Sec. 72.3.
    The definition for Transport index (TI) would be revised to reflect 
the new definition of Criticality Safety Index; however, the method for 
determining the TI of a package, based on the package's radiation dose 
rate, would remain unchanged.
10 CFR 71.5  Transportation of Licensed Material
    No changes were made to the text of this section; however, it is 
included in the revision of this subpart for completeness.
10 CFR 71.6  Information Collection Requirements: OMB Approval
    This section would be redesignated from Subpart B--Exemptions, to 
Subpart A--General Provisions. Paragraph (b) of this section would be 
revised as a conforming change to reflect the addition of new 
information collection requirements in Secs. 71.18, 71.151, 71.153, 
71.155, 71.157, 71.159, 71.161, 71.165, 71.167, 71.171, 71.173, 71.175, 
and 71.177. Additionally, the existing information collection 
requirement in Appendix A to Part 71, Paragraph II, was inadvertently 
omitted from the list of approved information collection requirements 
in a previous rulemaking; consequently, NRC staff would add Appendix A, 
Paragraph II, to paragraph (b) to correct this error. Furthermore, 
Sec. 71.6a would be removed, because no such section currently exists 
in Part 71.
10 CFR 71.7  Completeness and Accuracy of Information
    This section would be redesignated from Subpart B--Exemptions, to 
Subpart A--General Provisions. Further, paragraphs (a) and (b) would be 
revised by adding the terms ``certificate holder'' and ``applicant for 
a CoC.''
10 CFR 71.8  Deliberate Misconduct
    This section would be redesignated from Subpart B--Exemptions, to 
Subpart A--General Provisions. Further, in Subpart A, Sec. 71.11 would 
be redesignated as Sec. 71.8. However, the current text of Sec. 71.11 
would not be changed in the redesignated Sec. 71.8.
10 CFR 71.9  Employee Protection
    New Sec. 71.9 would be added to provide requirements on employee 
protection. Currently, requirements relating to the protection of 
employees against firing or other discrimination when the employee 
engages in certain ``protected activities'' are provided under the 
Parts of Title 10 for which a specific license was issued to possess 
radioactive material. However, no provisions were provided in Part 71 
relating to the protection of employees against firing or other 
discrimination when employees engage in certain ``protected 
activities'' when they are the employees of a certificate holder or 
applicant for a CoC. The NRC believes these employees should also be 
afforded the same rights and protection as are currently afforded 
employees of licensees. The new section would be identical to the 
existing Sec. 72.10, ``Employee protection.'' In including licensees in 
the new Sec. 71.9, the NRC recognizes that the potential for 
duplication occurs for licensees regulated under multiple 10 CFR Parts. 
However, the NRC believes that by including licensees along with 
certificate holders and applicants for a CoC, improved regulatory 
clarity would be achieved, and any potential confusion would be 
minimized.
10 CFR 71.10  Public Inspection of Application
    A new section would be added indicating that applications and 
documents submitted to the Commission in connection with an application 
for a package approval shall be available for public review in 
accordance with the provisions of 10 CFR Parts 2 and 9. This new 
section would be similar to existing Sec. 72.20. Existing Sec. 71.10 
would be redesignated Sec. 71.14 with changes to the text.
10 CFR 71.11  (Reserved)
    This section would be redesignated from Subpart B--Exemptions, to 
Subpart A--General Provisions, and would be reserved. Existing 
Sec. 71.11 would be redesignated as Sec. 71.8.

Subpart B--Exemptions

10 CFR 71.12  Specific Exemptions
    Existing Sec. 71.8 would be redesignated as Sec. 71.12. No changes 
would be made to the contents of this section. Existing Sec. 71.12 
would be redesignated as Sec. 71.17, with changes to the text as 
discussed under Sec. 71.17, below.
10 CFR 71.13  Exemption of Physicians
    Existing Sec. 71.9 would be redesignated as Sec. 71.13. No changes 
would be made to the contents of this section. Existing Sec. 71.13 
would be redesignated as Sec. 71.19, with changes to the text as 
discussed under Sec. 71.19, below.

[[Page 21430]]

10 CFR 71.14  Exemption for Low-Level Materials
    Existing Sec. 71.10 would be redesignated as Sec. 71.14. Existing 
Sec. 71.14 would be redesignated as Sec. 71.20, with no changes to the 
text.
    In new Sec. 71.14, paragraph (a) would be revised by removing the 
existing single 70 Bq/g (0.002 Ci/g) specific activity value 
and replacing it with ``Activity Concentration for Exempt Material'' 
found in Table A-2 in Appendix A to Part 71. Additionally, paragraph 
(a) would be reformatted by adding two new paragraphs. Paragraph (a)(1) 
would provide an exemption for natural radioactive materials and ores. 
Paragraph (a)(2) would provide an exemption for radioactive material 
based on its specific activity, not based on the material being in a 
package.
    Paragraph (b) would be revised to consolidate the exemption 
provisions for LSA and SCO material. The LSA and SCO exemptions 
contained in existing paragraphs (b)(2) and (c) of this section would 
be consolidated into a revised paragraph (b)(3), and existing paragraph 
(c) would be removed. The reference to material exempt from 
classification as fissile material would be revised from Sec. 71.53 to 
Sec. 71.15, because of the redesignation of the section.
    Existing paragraph (b)(3) would be removed. The 0.74-TBq (20-Ci) 
exemption for special form americium and special form plutonium would 
be removed. However, the 0.74-TBq (20-Ci) exemption for special form 
plutonium-244, transported in domestic commerce, would be retained as 
new paragraph (b)(2). Furthermore, an exception would be added to 
paragraph (b)(1) indicating that paragraph (b)(1) does not apply to a 
package containing greater than an A1 quantity of special form 
plutonium-244 transported in domestic commerce. For international 
shipments, the A1 quantity limit for special form plutonium-244 would 
continue to apply.
10 CFR 71.15  Exemption From Classification as Fissile Material
    Existing Sec. 71.11 would be redesignated to Sec. 71.8. Existing 
Sec. 71.53 would be redesignated as Sec. 71.15, and relocated to 
Subpart B with the other Part 71 exemptions. This section would be 
revised by providing mass-ratio based limits in classifying fissile-
exempt material. This approach would remove the concentration- and 
consignment-based limits of the current Sec. 71.53 and return to 
package-based mass limits, with required minimum ratios of nonfissile-
to-fissile mass.
    The title would be changed to ``Exemption from classification as 
fissile material.''
    New paragraphs (a) and (b) would be added and would allow for 
increasing quantities of fissile material to be shipped, would provide 
a concurrent increase in the required mass ratio to ensure criticality 
safety, and would allow shipment of fissile material in bulk packaging 
(i.e., large freight containers). The nonfissile material would be 
limited to noncombustible material which is insoluble in water. In 
paragraph (a), the fissile mass per package would be limited to 15 
grams with a nonfissile-to-fissile mass ratio of 200:1, and the 
nonfissile material would be restricted to iron. In paragraph (b), the 
allowed fissile mass is raised to 350 grams per package, but the ratio 
of nonfissile-to-fissile material is also raised to 2000:1. The mass of 
any lead, graphite, beryllium, and deuterium in the package cannot be 
included in determining the nonfissile material mass, and the 
nonfissile material that is counted in the ratio must be noncombustible 
and insoluble in water.
    Current Sec. 71.53, paragraph (b), would be redesignated as 
paragraph (c), and would be revised to limit beryllium, graphite, and 
hydrogenous material enriched in deuterium to less than 0.1 percent of 
the fissile material mass. The current homogenous distribution and 
lattice requirements would be removed.
    Current Sec. 71.53, paragraph (c), would be redesignated as 
paragraph (d), and would be reformatted and revised to clarify that the 
nitrogen to uranium atomic ratio, for shipments of liquid uranyl 
nitrate, must be greater than or equal to 2.0. A new requirement would 
be added specifying the use of DOT Type A packaging.
    Current Sec. 71.53, paragraph (d), would be redesignated as 
paragraph (e), and would be reformatted and revised to clarify the mass 
limits for plutonium. No substantive changes would be made to this 
paragraph.
10 CFR 71.16  (Reserved)
    This section would be redesignated from Subpart C--General 
Licenses, to Subpart B--Exemptions, and would be reserved. Further, 
existing Sec. 71.16 would be redesignated as Sec. 71.21. However, the 
current text of Sec. 71.16 would not be changed in the redesignated 
Sec. 71.21.

Subpart C--General Licenses

Section 71.17  General license: NRC-Approved package
    Existing Sec. 71.12 would be redesignated as Sec. 71.17. Paragraph 
(a) would be revised as a conforming change to indicate that this 
general license does not apply to Type B(DP) packages.
    Paragraph (c)(3) would be revised using plain language, and to 
reflect the NRC's requirement to address information submitted to the 
NRC to the attention of the NRC's Document Control Desk, in accordance 
with Sec. 71.1.
10 CFR 71.18  General License: NRC-Approved Type B(DP) Package
    This new section would be added to provide a general license for 
the transportation of spent fuel in Type B(DP) packages. The structure 
of this new section would be similar to the existing Sec. 71.12(a) 
through (d).
10 CFR 71.19  Previously Approved Package
    Existing Sec. 71.13 would be redesignated as Sec. 71.19. Paragraph 
(a) would be revised to reflect the current package designators (e.g., 
B(U)F, B(M)F, AF). Additionally, the contents of paragraph (a)(2) would 
be removed to reflect that these packages are no longer recognized 
internationally. Existing paragraph (a)(3) would be redesignated as 
(a)(2) with no change to the contents. Also, an expiration date for 
grandfathering these packages would be established. Paragraph (b) would 
be updated to remove the LSA packages, as these packages no longer 
exist. A new paragraph (c) would be added to reflect the type B(U) and 
B(M) packages that have met the requirements of IAEA Safety Series 6 
1985 (as amended 1990). Additionally, a date by which fabrication of 
these packages must be complete would be added. Existing paragraph (c) 
would be redesignated as paragraph (d). Existing paragraph (d) would be 
redesignated as paragraph (e), and updated to reflect the 
identification number suffix of ``-96'' for previously approved package 
designs that have been resubmitted for review by the NRC and have been 
approved, and to remove the package designated as Type A from this 
paragraph.
10 CFR 71.20  General License: DOT Specification Container
    Existing Sec. 71.14 would be redesignated as Sec. 71.20. No changes 
would be made to the contents of this section.
10 CFR 71.21  General License: Use of Foreign Approved Package
    Existing Sec. 71.16 would be redesignated as Sec. 71.21. No changes 
would be made to the contents of this section.

[[Page 21431]]

10 CFR 71.22  General License: Fissile Material
    Existing Sec. 71.18 would be redesignated as Sec. 71.22. This 
section would be amended by consolidating and simplifying the current 
fissile general license provisions contained in existing Secs. 71.18, 
71.20, 71.22, and 71.24 into a new Sec. 71.22. The new Sec. 71.22, 
while retaining some of the provisions of the existing general 
licenses, would principally use mass-based limits and a CSI. 
Concentration-based limits would be removed. Exceptions relating to 
plutonium-beryllium sealed sources in existing Secs. 71.18 and 71.22 
would be relocated to new Sec. 71.23. The values contained in new 
Tables 71-1 and 71-2 would be revised from the values contained in the 
table in existing Sec. 71.22 and in Table 1 in existing Sec. 71.20, 
respectively; and are based on new minimum critical mass calculations 
described in NUREG/CR-5342. In some instances, the allowable mass limit 
has been increased from the current limits in existing Secs. 71.18, 
71.20, 71.22, and 71.24; in other instances, the allowable mass limit 
has been reduced. The values contained in new Tables 71-1 and 71-2 
would be used as the variables X, Y, and Z in the equation in paragraph 
(e).
    The title would be revised to indicate that this general license is 
not restricted to a specific type of fissile material shipment.
    Paragraph (a) would be revised to require that fissile material 
shipped under this general license would be contained in a DOT Type A 
package. Additionally, while the existing exception from Subparts E and 
F requirements is maintained, the DOT Type A package regulations of 49 
CFR Part 173 would also be specified.
    Paragraph (b) would remain unchanged.
    Paragraph (c) would be revised to remove the specific gram limits 
for uranium and plutonium, but would retain the existing Type A 
quantity limit. Revised gram limits would be relocated to new Table 71-
1, which would be associated with new paragraphs (d) and (e). A 
requirement would also be added to limit the amount of special 
moderating materials beryllium, graphite, and hydrogenous material 
enriched in deuterium present in a package to less than 500 g.
    Existing paragraph (d) would be removed. Revised gram limits for 
fissile material mixed with material having a hydrogen density greater 
than water (i.e., a moderating effectiveness greater than 
H2O) would be placed in new Table 71-1. A note would be 
added to new Table 71-1 to indicate that reduced mass limits apply when 
more than 15 percent of a mixture of moderating materials contains 
moderating material with a hydrogen density greater than 
H2O.
    New paragraph (d) would be added to require that shipments of 
packages containing fissile material be labeled with a CSI, that the 
CSI per package be less than or equal to 10.0, and that the sum of the 
CSIs in a shipment of multiple fissile material packages would be 
limited to less than or equal to 50.0 for a nonexclusive use 
conveyance, and to less than or equal to 100.0 for an exclusive use 
conveyance.
    Existing paragraphs (e) and (f) would be removed.
    New paragraph (e) would be added to require that the CSI be 
calculated via a new equation for any of the fissile nuclides. Guidance 
on applying the equation and the mass limit input values of Tables 71-1 
and 71-2 would also be contained in this paragraph.
10 CFR 71.23  General License: Plutonium-Beryllium Special Form 
Material
    The existing Sec. 71.20, ``General license: Fissile material, 
limited moderator per package,'' would be removed. A new section on the 
shipment of plutonium-beryllium (Pu-Be) special-form fissile material 
(i.e., sealed sources) would be added as a new Sec. 71.23. New 
Sec. 71.23 would consolidate regulations on shipment of Pu-Be sealed 
sources contained in existing Secs. 71.18 and 71.22 into one location 
in Part 71 and would use an approach consistent with the revised 
Sec. 71.18. The Sec. 71.23 would reduce the maximum quantity of fissile 
plutonium Pu-Be sealed sources that could be shipped on a single 
conveyance through changes in the mass limits and calculation of the 
CSI. Currently, a Pu-Be sealed source package can contain up to 400 g 
of fissile plutonium with a CSI equal to 10.0. Consequently, the 
current conveyance limits are 4,000 g per shipment for an exclusive-use 
vehicle and 2000 g per shipment for a nonexclusive use vehicle. The new 
Sec. 71.23 would increase the maximum CSI per package from 10 to 100; 
however, the maximum quantity of plutonium per conveyance (i.e., 
shipment) would be reduced to 1000 g. The 1000 g per shipment limit and 
a 240 g of fissile plutonium limit are equivalent to those in new 
Sec. 71.22(f) (1,000 g per shipment and 200 g of fissile plutonium). 
The 240 g versus 200 g of fissile plutonium per package is due to the 
increased confidence that the fissile plutonium within a sealed source 
capsule would not escape from the capsule during an accident and 
reconfigure itself into an unfavorable geometry.
    New Sec. 71.23 would be titled: ``General license: Plutonium-
beryllium special form material.''
    Paragraph (a) would describe the applicability of this section, 
exceptions to the requirements of Subparts E and F, and the requirement 
to ship Pu-Be sealed sources in DOT Type A packages.
    Paragraph (b) would require that shipments of Pu-Be sealed sources 
be made under an NRC-approved QA program.
    Paragraph (c) would require a 1,000 g per package limit. In 
addition, plutonium-239 and plutonium-241 may constitute only 240 g of 
the 1,000 g limit.
    Paragraph (d) would require that a CSI be calculated per paragraph 
(e), and the CSI must be less than or equal to 100.0. For shipments of 
multiple packages, the sum of the CSIs would be limited to less than or 
equal to 50.0 for a nonexclusive use conveyance, and to less than or 
equal to 100.0 for an exclusive use conveyance.
    Paragraph (e) would provide an equation to calculate the CSI for 
Pu-Be sources. This equation would be based upon the 240 g mass limit 
for fissile nuclide plutonium-239 and plutonium-241 in paragraph (c).
10 CFR 71.24  (Reserved)
10 CFR 71.25  (Reserved)
    Existing Secs. 71.22 and 71.24 would be redesignated as Secs. 71.24 
and 71.25. New Secs. 71.24 and 71.25 would be removed and reserved.

Subpart D--Application for Package Approval

10 CFR 71.41  Demonstration of Compliance
    Paragraph (a) would be revised to require that a Type B package 
which contains radioactive contents with activity greater than 10 
5 A 2 of any radionuclide must meet the enhanced 
deep immersion test found in Sec. 71.61. A new paragraph (d) would be 
added to provide special package authorizations.
10 CFR 71.51  Additional Requirements for Type B Packages
    Paragraph (a) would be revised to remove the reference to 
Sec. 71.52, because the requirements of Sec. 71.52 have expired. 
Paragraph (d) would be added to require that, for other than Type B(DP) 
packages, a package which contains radioactive contents with activity 
greater than 10 5 A 2 of any radionuclide must 
also meet the

[[Page 21432]]

enhanced deep immersion test found in Sec. 71.61.
10 CFR 71.53  Fissile Material Exemptions (Reserved)
    This section would be removed and reserved; its contents would be 
moved to Sec. 71.15.
10 CFR 71.55  General Requirements for Fissile Material Packages
    New paragraphs (f) and (g) would be added. Paragraph (f) would 
specify design and testing for fissile material package design for 
transport by aircraft, and paragraph (g) would address UF6 
criticality exception from Sec. 71.55(b). Additionally, as a conforming 
change, paragraph (b) would be updated to support new paragraph (g).
10 CFR 71.59  Standards for Arrays of Fissile Material Packages
    Paragraphs (b) and (c) would be revised to use the term CSI 
(criticality safety index).
    Paragraph (b) would be revised to refer to a CSI rather than a TI 
for nuclear criticality control. The method for calculating a CSI would 
be the same as the existing method for a TI for nuclear criticality 
control.
    Paragraph (c) of this section would be revised to provide direction 
to licensees when the CSI is exactly equal to 50.0, and to use plain 
language. Subparagraph (1) Would be revised by replacing the term 
``[n]ot in excess of 10,'' with the term ``[l]ess than or equal to 
50.0,'' and would provide for storage incident to transport. New 
paragraph (c)(2) would be added to provide for shipment of packages 
with a CSI of less than 50.0 on an exclusive use conveyance. The 
current conveyance limit of 100 would be retained. Existing paragraph 
(c)(2) would be redesignated as new paragraph (c)(3) and would be 
revised by replacing the term ``[i]n excess of 10,'' with the term 
``[g]reater than 50.0.'' These three changes would: (1) Provide greater 
clarity and mathematical consistency among paragraphs (c)(1), (c)(2), 
and (c)(3); (2) clarify the CSI limits for storage incident to 
transport; and (3) increase the CSI limit per package from 10 to 50 for 
shipments made with nonexclusive use conveyances.
10 CFR 71.61  Special Requirements for Type B Packages Containing More 
Than 10\5\ A2
    This section would be revised to require an enhanced water 
immersion test for packages used for radioactive contents with activity 
greater than 10\5\ A2. The title of this section would also 
be revised to reflect that the scope has been broadened beyond 
irradiated nuclear fuel.
10 CFR 71.63  Special Requirement for Plutonium Shipments
    The title would be revised to reflect only a single ``requirement'' 
rather than multiple requirements.
    Paragraph (b) would be removed.
    The designation of the remaining text as paragraph (a) would be 
removed, because only one paragraph would remain. The text of former 
paragraph (a) would be revised to use plain language. The 0.74-TBq (20-
Ci) limit and solid form requirement would be retained.
10 CFR 71.73  Hypothetical Accident Conditions
    A new paragraph (c)(2) is added to require a crush test for fissile 
material packages.
10 CFR 71.88  Air Transport of Plutonium
    Paragraph (a)(2) would be revised to remove the 70-Bq/g (0.002-
Ci/g) specific activity value and substitute activity 
concentration values for plutonium found in Appendix A, Table A-2, of 
this part. This revision would be a conforming change to the revision 
to new Sec. 71.14 to ensure consistent treatment of plutonium between 
these two sections.

Subpart G--Operating Controls and Procedures

10 CFR 71.91  Records
    As a conforming change to Subpart H, paragraphs (b) and (c) would 
be redesignated as paragraphs (c) and (d), respectively, and would be 
revised by adding the terms certificate holder and applicant for a CoC. 
New paragraph (b) would be added to require a certificate holder to 
keep records on the model, serial number, and date of manufacture of a 
packaging. These requirements are similar to the requirements in 
paragraph (a), though less information is required. No change would be 
made to paragraph (a).
10 CFR 71.93  Inspection and Tests
    As a conforming change to Subpart H, paragraphs (a) and (b) would 
be revised by adding the terms certificate holder and applicant for a 
CoC. Paragraph (c) would be revised to require the certificate holder 
to notify the NRC before it begins fabrication of a packaging that can 
contain material having a decay heat load in excess of 5 kW or a 
maximum normal operating pressure of 103 kPa [kilo Pascals] (15 lbf/in 
\2\) gauge. This notification could be for either fabricating a single 
packaging or the beginning of a campaign for fabricating multiple 
packagings. This notification would be in accordance with the 
requirements of Sec. 71.1, rather than to an NRC Regional 
Administrator. This change in notification location is consistent with 
current Commission policy and would reduce confusion in identifying the 
appropriate Regional Administrator when the certificate holder and 
fabrication location are overseas. Licensees would be removed from this 
paragraph because the NRC believes that requiring a licensee, who does 
not own the packaging, to notify the NRC in advance of a packaging 
fabrication, when the licensee may not use the packaging for years, is 
inappropriate and an unreasonable burden. The NRC believes that 
requiring certificate holders and applicants for a CoC to notify the 
NRC in advance of fabricating a packaging(s) would allow the NRC 
adequate opportunity to inspect these activities. This change would be 
similar to the current requirement in Sec. 72.232(d) for Part 72 
certificate holders or applicants for a CoC to notify the NRC 45 days 
before starting the fabrication of the first storage cask under a part 
72 CoC. This action would improve the harmonization between these two 
regulations in parts 71 and 72, particularly regarding dual-purpose 
casks (i.e., casks intended to both store and transport spent fuel).
10 CFR 71.95  Reports
    The existing introductory text and paragraphs (a) and (b) would be 
combined into a new paragraph (a) which would require a licensee, after 
requesting the certificate holder's input, to submit a written report 
to the NRC in certain circumstances. The requirement for the licensee 
to request input from the certificate holder during development of the 
written event report would ensure that design deficiency issues have 
been thoroughly considered. The licensee would also be required to 
provide the certificate holder with a copy of the written event report, 
after the report is submitted to the NRC. This would permit the 
certificate holder to monitor and trend the package performance 
information, arising from package use by multiple licensees. 
Additionally, requirements on timing and submission location for the 
written reports would be relocated to new paragraph (c). Furthermore, 
the 30-day reporting requirement would be lengthened to a 60-day 
reporting requirement.
    The existing paragraph (c) has been redesignated as paragraph (b) 
and revised for clarity.

[[Page 21433]]

    New paragraphs (c) and (d) would be added to provide requirements 
on the timing, submission location, form, and content of the written 
reports.
10 CFR 71.100  Criminal Penalties
    Section 223 of the Atomic Energy Act of 1954, as amended, [the Act] 
provides for criminal sanctions for willful violation of, attempted 
violation of, or conspiracy to violate, any regulation issued under 
sections 161b, 161i, or 161o of the Act. The Commission stated in a 
final rule on ``Clarification of Statutory Authority for Purposes of 
Criminal Enforcement'' (57 FR 55082; November, 24, 1992), that 
substantive rules under sections 161b, 161i, or 161o of the Act include 
those rules that create ``duties, obligations, conditions, 
restrictions, limitations, and prohibitions.'' For the NRC to consider 
the possibility of criminal sanctions for willful violation of, 
attempted violation of, or conspiracy to violate, any substantive 
regulations, the NRC must have clearly identified to affected parties 
which regulations in Part 71 are substantive rules. Accordingly, 
paragraph (b) of this section identifies those Part 71 regulations that 
the NRC does not consider as substantive regulations. Thus, willful 
violation of, attempted violation of, or conspiracy to violate any of 
the regulations listed in paragraph (b) is not subject to possible 
criminal sanctions.
    Paragraph (b) of this section would be revised as a conforming 
change. The NRC has reviewed new Secs. 71.10, 71.151, 71.153, 71.155, 
71.157, 71.159, 71.161, 71.163, 71.165, 71.167, and 71.169 and 
considers that these regulations are not substantive rules. Therefore, 
new Secs. 71.10 and 71.151 through 71.169 would be added to the list of 
sections in paragraph (b). The NRC reviewed new Secs. 71.9, 71.18, 
71.23, 71.171, 71.173, 71.175, and 71.177, and considers that these 
regulations are substantive rules. Therefore, these sections would not 
be added to paragraph (b). Additionally, the NRC has reviewed the 
existing Secs. 71.9, 71.10, and 71.53 and concluded these sections 
should be recharacterized as substantive rules. Therefore, new 
Secs. 71.13, 71.14, and 71.18 would not be included in paragraph (b). 
Additionally, existing Secs. 71.52 and 71.53 would be removed from 
paragraph (b), because these section numbers have been removed from 
part 71.

Subpart H--Quality Assurance

10 CFR 71.101  Quality Assurance Requirements
    Paragraph (a) would be revised by adding two new sentences to the 
end of the paragraph specifying responsibilities for certificate 
holders and applicants for a CoC.
    Paragraph (b) would be revised to add the terms ``certificate 
holder'' and ``applicant for a CoC.'' The second sentence would be 
revised to provide greater clarity and consistency within Subpart H by 
referring to ``the QA requirement's importance to safety.''
    Paragraph (c) would be revised by redesignating the existing text 
as paragraph (c)(1), and new text would be added on submitting QA 
programs in accordance with the requirements of Sec. 71.1. New 
paragraph (c)(2) would be added to provide equivalent requirements on 
the submission of QA programs for certificate holders and applicants 
for a CoC.
    Paragraph (f) would be revised to allow the use of existing NRC-
approved Part 71 and Part 72 QA programs, in lieu of submitting a new 
QA program. Additionally, the terms ``certificate holder'' and 
``applicant for a CoC'' would be added.
    Paragraph (g) would be revised by making a minor change to clarify 
that Sec. 34.31(b) is located in Chapter I of Title 10 of the Code of 
Federal Regulations. Additionally, as a conforming change, 
Sec. 71.12(b) would be redesignated as Sec. 71.17(b).
10 CFR 71.103  Quality Assurance Organization
    Paragraph (a) would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.'' Further, the fourth sentence 
would be revised to improve clarity and consistency within Subpart H 
and with Part 72, Subpart G, by referring to ``the functions of 
structures, systems, and components that are important to safety.''
10 CFR 71.105  Quality Assurance Program
    Paragraphs (a) through (d) would be revised by adding the terms 
``certificate holder'' and ``applicant for a CoC.''
10 CFR 71.107  Package Design Control
    Paragraph (a) would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.'' Further, the last sentence would 
be revised to improve clarity and consistency within Subpart H by 
referring to ``processes that are essential to the functions of the 
materials, parts, and components that are important to safety.''
    Paragraph (b) would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.'' Additionally, the last sentence 
would be revised by replacing the text ``[c]hanges in the conditions 
specified in the package approval require NRC approval. * * *'' with 
``[c]hanges in the conditions specified in the CoC require NRC prior 
approval. * * *''
10 CFR 71.109  Procurement Document Control
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.111  Instructions, Procedures, and Drawings
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.113  Document Control
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.115  Control of Purchased Material, Equipment, and Services
    Paragraphs (a) through (c) would be revised by adding the terms 
``certificate holder'' and ``applicant for a CoC.''
10 CFR 71.117  Identification and Control of Materials, Parts, and 
Components
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.119  Control of Special Processes
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.121  Internal Inspection
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.123  Test Control
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.125  Control of Measuring and Test Equipment
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.127  Handling, Storage, and Shipping Control
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.129  Inspection, Test, and Operating Status
    Paragraph (a) would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
    Paragraph (b) would remain unchanged.

[[Page 21434]]

10 CFR 71.131  Nonconforming Materials, Parts, or Components
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.133  Corrective Action
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.135  Quality Assurance Records
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''
10 CFR 71.137  Audits
    This section would be revised by adding the terms ``certificate 
holder'' and ``applicant for a CoC.''

Subpart I--Application for Type B(DP) Package Approval

    New Subpart I would be added to provide requirements on the 
application, review, approval, and amendment of a CoC for a Type B(DP) 
package. Requirements would also be provided on the submission and 
periodic updating of an FSAR. Additionally, requirements would be added 
authorizing a certificate holder to make minor changes to the design of 
a Type B(DP) package, without prior NRC approval, if certain tests were 
met. Further, identification would be made of which sections in Part 71 
also apply to packages approved under this new subpart.
10 CFR 71.151  Procedures for Applying for a Type B(DP) Package 
Approval
    This new section would describe the process for submitting an 
application to the NRC to request approval of a Type B(DP) package 
design. This section would be similar to Sec. 72.230.
10 CFR 71.153  Contents of Application
    This new section would provide requirements on what information 
must be contained in an application for a Type B(DP) package approval. 
This section would be similar to Sec. 71.31.
10 CFR 71.155  Package Description
    This new section would provide requirements on the description of a 
Type B(DP) package (both the packaging and its contents) which must be 
contained in an application for package approval. This section would be 
similar to Sec. 71.33.
10 CFR 71.157  Package Evaluation
    This new section would provide requirements which an application 
for a Type B(DP) package must demonstrate compliance with (i.e., 
sections in Subparts E and F). Additionally, because the Type B(DP) 
package is a fissile material package, the applicant would be required 
to: (1) Determine and provide the number ``N'' which is used in 
determining the maximum number of fissile packages on a conveyance; and 
(2) provide any special controls, precautions, or handling 
instructions. This section would be similar to Sec. 71.35.
10 CFR 71.159  Quality Assurance
    This new section would require a certificate holder to describe the 
quality assurance program, which meets the requirements of Subpart H of 
Part 71, that would be used to design, fabricate, test, repair, and 
modify a Type B(DP) package. This section would be similar to 
Sec. 71.37.
10 CFR 71.161  Requirement for Additional Information
    This new section would require a certificate holder to provide the 
Commission any information the NRC requires to determine if a CoC 
should be modified, suspended, or revoked. This section would be 
similar to Sec. 71.39.
10 CFR 71.163  Issuance of an NRC Certificate of Compliance
    This new section would provide direction to the NRC staff on 
criteria for approving a Type B(DP) CoC. This section would be similar 
to Sec. 72.238.
10 CFR 71.165  Conditions for Package Reapproval
    This new section would provide direction to a certificate holder 
who desires to renew a Type B(DP) CoC or a Part 71 QA program approval. 
This section would be similar to Sec. 71.38.
10 CFR 71.167  Application To Amend a Certificate of Compliance
    This new section would provide direction to a certificate holder 
who wishes to amend the CoC for a Type B(DP) package. This section 
would be similar to Sec. 72.244.
10 CFR 71.169  Issuance of an Amendment to a Certificate of Compliance
    This new section would provide direction to the NRC staff on 
issuance of an amendment to a Type B(DP) package CoC. This section 
would be similar to Sec. 72.246.
10 CFR 71.171  Inspections and Tests
    This new section would require a certificate holder to permit and 
to make provisions for NRC inspections at facilities used to design, 
fabricate, or test a Type B(DP) package. This section would also 
require a certificate holder to make records available and to perform 
tests the Commission deems necessary. This section would be similar to 
Sec. 72.232.
10 CFR 71.173  Recordkeeping and Reports
    This new section would provide requirements on submitting reports 
to the NRC and on maintaining records of fabricated Type B(DP) 
packages. This section would be similar to Sec. 72.242.
10 CFR 71.175  Changes
    This new section would provide requirements permitting a Part 71 
certificate holder to make changes to the design of a Type B(DP) 
package without prior NRC approval. The certificate holder would be 
required to periodically submit to the NRC a summary of any changes 
made under Sec. 71.175. This section would be similar to Sec. 72.48.
10 CFR 71.177  Safety Analysis Report Updating
    This new section would provide requirements for a Type B(DP) 
certificate holder on: (1) An initial submittal of an FSAR to the NRC; 
(2) submitting periodic updates of the FSAR to the NRC; and (3) 
providing a copy of the updated FSAR to each licensee using the Type 
B(DP) package. This section would be similar to Sec. 72.248.

Appendix A to Part 71--Determination of A1 and 
A2

    No changes were made in Paragraphs I, III, and V; however, these 
paragraphs would be included due to revising Appendix A in its 
entirety.
    Paragraph II would be revised to use plain language and would be 
redesignated as subparagraph II(a). The intent of existing paragraph II 
would not be changed; however, the reference to existing Table A-2 
would be revised as a conforming change to the new Table A-3. New 
paragraph II(b) would be added to provide direction on determining 
exempt material activity concentration and exempt consignment activity 
values when a radionuclide has been identified as a constituent of a 
proposed shipment, but the individual radionuclide is not listed in 
Table A-2. Consequently, the structure of paragraphs II(a) and II(b) 
would be the same. New paragraph II(c) would be added to provide 
direction to licensees on how to submit requests for Commission prior 
approval of either A1 and A2 values or exempt 
material activity concentration and exempt

[[Page 21435]]

consignment activity values, for radionuclides that are not listed in 
Tables A-1 and A-2, respectively.
    Paragraph IV would be revised by adding new paragraphs (e) and (f) 
to provide equations to use in determining a consolidated exempt 
material activity concentration and exempt consignment activity values 
when a shipment contains multiple radionuclides. The existing text 
describing an alternative method for calculating the A1 and 
A2 value of a mixture would be redesignated as paragraphs 
(c) and (d). No changes would be made from the existing equations.

Appendix A, Table A-1--A1 and A2 Values for 
Radionuclides

    This Table would be revised to reflect the values from TS-R-1.
Appendix A, Table A-2--Exempt Material Activity Concentrations and 
Exempt Consignment Activity Limits for Radionuclides
    A new Table A-2 would be added to Appendix A of Part 71. This table 
would contain the values of Exempt Material Activity Concentrations and 
Exempt Consignment Activity Limits for selected radionuclides. Table A-
2 is referenced in new Sec. 71.14(a)(2), and is used by Sec. 71.14 to 
determine when concentrations of material are not considered 
radioactive material, for the purposes of transportation.
Appendix A, Table A-3--General Values for A1 and 
A2
    The existing Table A-2 would be redesignated as new Table A-3, and 
the values would be revised to reflect the changes from IAEA TS-R-1.
Appendix A, Table A-4--Activity Mass Relationships for Uranium
    The existing Table A-3 would be redesignated as new Table A-4. No 
changes would be made to the values contained in new Table A-4.

VI. Criminal Penalties

    For the purposes of Section 223 of the Atomic Energy Act (AEA), the 
Commission is proposing to issue amendments to amend 10 CFR Part 71 
under one or more of sections 161b, 161i, or 161o of the AEA. Willful 
violations of the rule would be subject to criminal enforcement.
    The following is a list of substantive rule sections being revised 
or added in this rulemaking: Secs. 71.1, 71.3, 71.5, 71.8, 71.9, 71.12, 
71.13, 71.14, 71.15, 71.17, 71.18, 71.19, 71.20, 71.21, 71.22, 71.23, 
71.61, 71.63, 71.88, 71.91, 71.93, 71.95, 71.101, 71.103, 71.105, 
71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 
71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, 71.171, 71.173, 
71.175, and 71.177.

VII. Issues of Compatibility for Agreement States

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' which became effective on September 3, 1997 
(62 FR 46517), NRC program elements (including regulations) are placed 
into four compatibility categories. In addition, NRC program elements 
also are identified as having particular health and safety significance 
or as being reserved solely to the NRC. Compatibility Category A are 
those program elements that are basic radiation protection standards 
and scientific terms and definitions that are necessary to understand 
radiation protection concepts. An Agreement State should adopt Category 
A program elements in an essentially identical manner in order to 
provide uniformity in the regulation of agreement material on a 
nationwide basis. Compatibility Category B are those program elements 
that apply to activities that have direct and significant effects in 
multiple jurisdictions. An Agreement State should adopt Category B 
program elements in an essentially identical manner. Compatibility 
Category C are those program elements that do not meet the criteria of 
Category A or B, but the essential objectives of which an Agreement 
State should adopt to avoid conflict, duplication, gaps, or other 
conditions that would jeopardize an orderly pattern in the regulation 
of agreement material on a nationwide basis. An Agreement State should 
adopt the essential objectives of the Category C program elements. 
Compatibility Category D are those program elements that do not meet 
any of the criteria of Category A, B, or C, above, and, thus, do not 
need to be adopted by Agreement States for purposes of compatibility. A 
bracket around a category means that the section may have been adopted 
elsewhere and it is not necessary to adopt it again. Health and Safety 
(H&S) are program elements that are not required for compatibility 
(i.e., Category D), but areidentified as having a particular health and 
safety role (i.e., adequacy) in the regulation of agreement material 
within the State. Although not required for compatibility, the State 
should adopt program elements in this category based on those of NRC 
that embody the essential objectives of the NRC program elements 
because of particular health and safety considerations. Compatibility 
Category NRC are those program elements that address areas of 
regulation that cannot be relinquished to Agreement States pursuant to 
the Atomic Energy Act, as amended, or provisions of Title 10 of the 
Code of Federal Regulations. These program elements should not be 
adopted by Agreement States. The following table lists the Part 71 
revisions and their corresponding categorization under the ``Policy 
Statement on Adequacy and Compatibility of Agreement State Programs.''

                         Part 71.--Packaging and Transportation of Radioactive Material
----------------------------------------------------------------------------------------------------------------
     Regulation  section             Section title        Compatibility category              Comments
----------------------------------------------------------------------------------------------------------------
Sec.  71.0...................  Purpose and Scope.......  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.1...................  Communications and        D.......................  This provision does not meet
                                Records.                                            any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.2...................  Interpretations.........  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.3...................  Requirements for license  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.4...................  Definitions.............                            .............................

[[Page 21436]]

 
                               A1......................  [B].....................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
                               A2......................  [B].....................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
                               Certificate of            D.......................  This definition is not
                                Compliance (CoC).                                   required for compatibility
                                                                                    since it defines a term
                                                                                    which pertains to an area
                                                                                    reserved to NRC. A State may
                                                                                    adopt this definition for
                                                                                    purposes of clarity or
                                                                                    communication. This
                                                                                    definition can be adopted by
                                                                                    Agreement States since it in
                                                                                    and of itself does not
                                                                                    convey any authority whereby
                                                                                    a State can regulate in an
                                                                                    exclusive NRC jurisdiction.
                                                                                    However, if a State chooses
                                                                                    to define the term then the
                                                                                    definition should be
                                                                                    essentially identical. In
                                                                                    addition, this term does not
                                                                                    meet any of the criteria of
                                                                                    the Category A, B, C, or
                                                                                    health and safety, and this
                                                                                    term is widely accepted as
                                                                                    an area of sole
                                                                                    responsibility of the NRC.
                               Criticality of Safety     B.......................  This definition is designated
                                Index.                                              Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. In addition, this
                                                                                    definition is needed for a
                                                                                    common understanding beyond
                                                                                    a plain dictionary meaning
                                                                                    of the term in order to
                                                                                    implement Secs.  71.22 and
                                                                                    71.23.
                               Deuterium...............  B.......................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. In addition, this
                                                                                    definition is needed for a
                                                                                    common understanding beyond
                                                                                    a plain dictionary meaning
                                                                                    of the term in order to
                                                                                    implement Sec.  71.15.
                               DOT.....................  D.......................  This term does not meet any
                                                                                    of the criteria of the
                                                                                    Category A, B, C, or health
                                                                                    and safety because it is a
                                                                                    widely accepted abbreviation
                                                                                    for the U. S. Department of
                                                                                    Transportation.
                               Fissile material........  [B].....................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
                               Graphite................  B.......................  This definition is needed for
                                                                                    a common understanding
                                                                                    beyond a plain dictionary
                                                                                    meaning of the term in order
                                                                                    to implement Sec.  71.15,
                                                                                    which has direct and
                                                                                    significant transboundary
                                                                                    effects.
                               High-level radioactive    D.......................  This definition is not
                                waste or HLW.                                       required for compatibility
                                                                                    since it defines a term
                                                                                    which pertains to an area
                                                                                    reserved to NRC. A State may
                                                                                    adopt this definition for
                                                                                    purposes of clarity or
                                                                                    communication. This
                                                                                    definition can be adopted by
                                                                                    Agreement States since it in
                                                                                    and of itself does not
                                                                                    convey any authority whereby
                                                                                    a State can regulate in an
                                                                                    exclusive NRC jurisdiction.
                                                                                    However, if a State chooses
                                                                                    to define the term, then the
                                                                                    definition should be
                                                                                    essentially identical.
                               Low Specific Activity     [B].....................  This definition is designated
                                (LSA) material.                                     Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
                               Package.................  [B].....................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.

[[Page 21437]]

 
                               Spent nuclear fuel or     D.......................  This definition is not
                                Spent fuel.                                         required for compatibility
                                                                                    since it defines a term
                                                                                    which pertains to an area
                                                                                    reserved to NRC. A State may
                                                                                    adopt this definition for
                                                                                    purposes of clarity or
                                                                                    communication. This
                                                                                    definition can be adopted by
                                                                                    Agreement States since it in
                                                                                    and of itself does not
                                                                                    convey any authority whereby
                                                                                    a State can regulate in an
                                                                                    exclusive NRC jurisdiction.
                                                                                    However, if a State chooses
                                                                                    to define the term, then the
                                                                                    definition should be
                                                                                    essentially identical.
                               Structures, systems, and  D.......................  This definition is not
                                components important to                             required for compatibility
                                safety (SSCs).                                      since it defines a term
                                                                                    which pertains to an area
                                                                                    reserved to NRC. A State may
                                                                                    adopt this definition for
                                                                                    purposes of clarity or
                                                                                    communication. This
                                                                                    definition can be adopted by
                                                                                    Agreement States since it in
                                                                                    and of itself does not
                                                                                    convey any authority whereby
                                                                                    a State can regulate in an
                                                                                    exclusive NRC jurisdiction.
                                                                                    However, if a State chooses
                                                                                    to define the term, then the
                                                                                    definition should be
                                                                                    essentially identical.
                               Transport index.........  [B].....................  This definition is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
Sec.  71.5...................  Transportation of         B.......................  This provision is designated
                                licensed material.                                  Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner.
Sec.  71.6...................  Information collection    D.......................  This provision does not meet
                                requirements: OMB                                   any of the criteria for
                                approval.                                           designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.7...................  Completeness and          D.......................  This provision does not meet
                                accuracy of information.                            any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.8...................  Deliberate misconduct...  C.......................  The Commission determined in
                                                                                    response to SECY-97-156 that
                                                                                    Agreement States should
                                                                                    adopt the essential
                                                                                    objectives of this
                                                                                    provision. If deliberate
                                                                                    misconduct and wrongdoing
                                                                                    issues involving Agreement
                                                                                    State licensees were not
                                                                                    pursued and closed by
                                                                                    Agreement States, then a
                                                                                    potential gap may be created
                                                                                    between NRC and Agreement
                                                                                    State programs.
Sec.  71.9...................  Employee protection.....  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.10..................  Public inspection of      D.......................  This provision does not meet
                                application.                                        any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.14..................  Exemptions for low level  B-paragraph (a) NRC-      Paragraph (a) is designated
                                material.                 paragraphs (b) and (c).   as a Compatibility Category
                                                                                    B because of the significant
                                                                                    transboundary impacts with
                                                                                    respect to the
                                                                                    implementation of the
                                                                                    ``Exempt Activity
                                                                                    Concentration Values,'' for
                                                                                    individual radionuclides in
                                                                                    Appendix A, which is
                                                                                    designated as a
                                                                                    Compatibility Category B.
                                                                                   Paragraphs (b) and (c) are
                                                                                    designated Compatibility
                                                                                    Category ``NRC.'' This
                                                                                    provision is reserved to the
                                                                                    NRC because it delineates
                                                                                    NRC's authority from that of
                                                                                    DOT's in the area of
                                                                                    transportation of
                                                                                    radioactive materials. These
                                                                                    provisions relinquish to DOT
                                                                                    the control of types of
                                                                                    shipment that are of low
                                                                                    risk both from radiation and
                                                                                    criticality standpoints.
                                                                                    Further, to ensure that only
                                                                                    low criticality risk
                                                                                    shipments are included in
                                                                                    the area of DOT authority,
                                                                                    these provisions restrict
                                                                                    the exemption to Type A and
                                                                                    Low-Specific-Activity (LSA)
                                                                                    or Surface Contaminated
                                                                                    Objects (SCOs) that either
                                                                                    contain no fissile material
                                                                                    or satisfy the fissile
                                                                                    material exemption
                                                                                    requirements in Sec.  71.15.
                                                                                    Finally, this provision is
                                                                                    reserved to the NRC because
                                                                                    this exemption does not
                                                                                    relieve licensees from DOT
                                                                                    requirements by reason of
                                                                                    NRC's authority. Thus,
                                                                                    Agreement States should not
                                                                                    adopt this provision in
                                                                                    order to retain their
                                                                                    ability to implement all of
                                                                                    49 CFR as directed by DOT.
Sec.  71.15..................  Exemptions from           [B].....................  This provision is designated
                                classification as                                   Compatibility Category B
                                fissile material.                                   because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.

[[Page 21438]]

 
Sec.  71.17..................  General license: NRC-     B.......................  This provision is designated
                                approved package.                                   Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner.
Sec.  71.18..................  General license: NRC-     NRC.....................  This provision is reserved to
                                approved Type B(DP)                                 NRC the NRC because it
                                package.                                            addresses packages intended
                                                                                    for both the storage and
                                                                                    transportation of spent
                                                                                    fuel.
Sec.  71.19..................  Previously approved       NRC.....................  This provision is reserved to
                                package.                                            the NRC because it addresses
                                                                                    packages intended for both
                                                                                    the storage and
                                                                                    transportation of spent
                                                                                    fuel.
Sec.  71.22..................  General license: Fissile  [B].....................  Sec.  71.22 was previously
                                material.                                           entitled, ``General license:
                                                                                    Fissile material, limited
                                                                                    quantity, controlled
                                                                                    shipment.'' It was
                                                                                    designated a Compatibility
                                                                                    Category D. As a part of
                                                                                    this amendment, this section
                                                                                    was removed.
                                                                                   This provision is designated
                                                                                    Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
Sec.  71.23..................  General license:          [B].....................  This provision is designated
                                Plutonium-beryllium                                 Compatibility Category B
                                special form material.                              because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner. The bracket ``B''
                                                                                    indicates that if a State
                                                                                    has adopted this definition
                                                                                    in another portion of its
                                                                                    regulations, such as the
                                                                                    State's DOT regulations,
                                                                                    then the adoption of this
                                                                                    definition is not necessary.
Sec.  71.24..................  [Reserved]..............  ........................  Sec.  71.24 was previously
                                                                                    entitled, ``General license:
                                                                                    Fissile material, limited
                                                                                    moderator, controlled
                                                                                    shipment.'' It was
                                                                                    designated a Compatibility
                                                                                    Category NRC. As a part of
                                                                                    this amendment, this section
                                                                                    was removed.
Sec.  71.25..................  [Reserved]..............  ........................  Sec.  71.25 is a new section
                                                                                    that is reserved.
Sec.  71.41..................  Demonstration of          NRC.....................  This provision is designated
                                compliance.                                         NRC because it addresses an
                                                                                    area reserved to NRC's
                                                                                    regulatory authority.
Sec.  71.51..................  Additional requirements   NRC.....................  This provision is designated
                                for Type B packages.                                NRC because it addresses an
                                                                                    area reserved to NRC's
                                                                                    regulatory authority, which
                                                                                    is the approval of Type B
                                                                                    packages.
Sec.  71.53..................  [Reserved]..............  ........................  Sec.  71.53 was previously
                                                                                    entitled, ``Fissile material
                                                                                    exemptions.'' It was
                                                                                    designated a Compatibility
                                                                                    Category NRC. As a part of
                                                                                    this amendment, the
                                                                                    provision was removed.
Sec.  71.55..................  General requirements for  NRC.....................   This provision is designated
                                fissile material                                    NRC because it addresses an
                                packages.                                           area reserved to NRC's
                                                                                    regulatory authority.
Sec.  71.59..................  Standards for arrays of   NRC.....................  This provision is designated
                                fissile material                                    NRC because it addresses an
                                packages.                                           area reserved to NRC's
                                                                                    regulatory authority.
Sec.  71.61..................  Special requirements for  NRC.....................  This provision is designated
                                Type B packages                                     NRC because it addresses an
                                containing morethan                                 area reserved to NRC's
                                10\5\A2.                                            regulatory authority.
Sec.  71.63..................  Special requirements for  NRC.....................  This provision is designated
                                plutonium shipments.                                NRC because it addresses an
                                                                                    area reserved to NRC's
                                                                                    regulatory authority.
Sec.  71.73..................  Hypothetical accident     NRC.....................  This provision is designated
                                conditions.                                         NRC because it addresses an
                                                                                    area reserved to NRC's
                                                                                    regulatory authority.
Sec.  71.88..................  Air transport of          B.......................  This provision is designated
                                plutonium.                                          Compatibility Category B
                                                                                    because it applies to
                                                                                    activities that have direct
                                                                                    and significant effects in
                                                                                    multiple jurisdictions. An
                                                                                    Agreement State should adopt
                                                                                    Category B program elements
                                                                                    in an essentially identical
                                                                                    manner.
Sec.  71.91..................  Records.................  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.93..................  Inspection and tests....  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.95..................  Reports.................  D.......................  This provision does not meet
                                                                                    any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.100.................  Criminal penalties......  D.......................  This provision does not meet
                                                                                    penalties any of the
                                                                                    criteria for designations
                                                                                    Category A, B, C, or health
                                                                                    and safety. Thus, it does
                                                                                    not need to be adopted by
                                                                                    Agreement States.

[[Page 21439]]

 
Sec.  71.101.................  Quality assurance         D--Paragraphs (a), (b),   Paragraphs (a), (b), and
                                requirements.             (c)(1) and (f) are        (c)(1) are designated
                                                          designated D for those    Category C and the essential
                                                          States which have no      objectives of these
                                                          licensees that use Type   provisions should be adopted
                                                          B packages.               by those Agreement States
                                                         C--Paragraphs (a), (b)     which have licensees who use
                                                          and (c)(1) are            Type B packages. These
                                                          designated C for those    provisions are designated
                                                          States which have         Category C because the
                                                          licensees that use Type   quality assurance of Type B
                                                          B packages..              packages is an activity that
                                                         D--paragraph (f)........   is needed in order to avoid
                                                         C--paragraph (g)........   a nationwide gap in the
                                                         NRC--paragraph (c)(2),     regulation of the
                                                          (d) and (e).              transportation of
                                                                                    radioactive materials. If
                                                                                    these provisions are not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objective of
                                                                                    paragraph (a) is that each
                                                                                    licensee who uses a Type B
                                                                                    package is responsible for
                                                                                    the quality assurance
                                                                                    requirements which apply to
                                                                                    the use of a package. The
                                                                                    essential objective of
                                                                                    paragraph (b) is that each
                                                                                    licensee who uses a Type B
                                                                                    package shall establish,
                                                                                    maintain, and execute a
                                                                                    quality assurance program.
                                                                                    The essential objective of
                                                                                    paragraph (c)(1) is that
                                                                                    each licensee who uses a
                                                                                    Type B package shall, prior
                                                                                    to the use of any package
                                                                                    for the shipment of any
                                                                                    material subject to this
                                                                                    part, obtain approval of its
                                                                                    quality assurance program by
                                                                                    the regulatory agency.
                                                                                   Paragraph (f) is not required
                                                                                    for compatibility because
                                                                                    the States have the
                                                                                    flexibility to determine
                                                                                    whether they wish to accept
                                                                                    a previously approved
                                                                                    quality assurance program.
Sec.  71.103.................  Quality assurance         D--for those States       For paragraph (a), those
                                organization.             which have no licensees   States which have licenses
                                                          that use Type B           that use Type B packages,
                                                          packages.                 and have adopted the
                                                         [C]--Paragraph (a) is      essential objectives of Sec.
                                                          designated.                71.101(a), it is not
                                                         [C] for those States       necessary for them to adopt
                                                          which have licensees      this provision again.
                                                          that use Type B          Paragraph (b) is designated
                                                          packages..                as a Category C and the
                                                         C--Paragraph (b) is        essential objectives of
                                                          designated C for those    these provisions should be
                                                          States which have         adopted by those Agreement
                                                          licensees that use Type   States which have licensees
                                                          B packages..              who use Type B packages.
                                                         D--paragraphs (d), (e),    This provision is designated
                                                          and (f).                  Category C because the
                                                                                    quality assurance of Type B
                                                                                    packages is an activity that
                                                                                    is needed in order to avoid
                                                                                    a nationwide gap in the
                                                                                    regulation of the
                                                                                    transportation of
                                                                                    radioactive materials. If
                                                                                    these provisions are not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objective of
                                                                                    paragraph (b) is that each
                                                                                    licensee who uses a Type B
                                                                                    package should verify by
                                                                                    procedures such as checking,
                                                                                    auditing, and inspection,
                                                                                    that activities affecting
                                                                                    the safety-related functions
                                                                                    have been performed
                                                                                    correctly.
Sec.  71.105.................  Quality assurance         D--for those States       Paragraphs (a) and (b) are
                                program.                  which have no licensees   designated Category C for
                                                          that use Type B           those States which have
                                                          packages.                 licensees that use Type B
                                                         C--Paragraphs (a) and      packages. These provisions
                                                          (b) are designated as C   are designated Category C
                                                          for those States which    because the quality
                                                          have licensees that use   assurance of Type B packages
                                                          Type B packages..         is an activity that is
                                                         D--paragraph (c)........   needed in order to avoid a
                                                                                    nationwide gap in the
                                                                                    regulation of the
                                                                                    transportation of
                                                                                    radioactive materials. If
                                                                                    these provisions are not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objectives of
                                                                                    paragraph (a) are that each
                                                                                    licensee who uses a Type B
                                                                                    package shall document the
                                                                                    quality assurance program by
                                                                                    written procedures or
                                                                                    instructions and shall carry
                                                                                    out the program in
                                                                                    accordance with those
                                                                                    procedures throughout the
                                                                                    period during which the
                                                                                    packaging is used, and shall
                                                                                    identify the material and
                                                                                    components covered by the
                                                                                    quality assurance program.
                                                                                    The essential objective of
                                                                                    paragraph (b) is that each
                                                                                    licensee who uses a Type B
                                                                                    package shall, through its
                                                                                    quality assurance program,
                                                                                    provide control over
                                                                                    activities affecting the
                                                                                    quality of the identified
                                                                                    materials and components to
                                                                                    an extent to assure that
                                                                                    Type B packages are shipped
                                                                                    and maintained in accordance
                                                                                    with the certificate of
                                                                                    compliance or other
                                                                                    approval.
Sec.  71.107.................  Package design control..  NRC.....................  This provision is reserved to
                                                                                    the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.109.................  Procurement document      NRC.....................  This provision is reserved to
                                control.                                            the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.111.................  Instructions,             NRC.....................  This provision is reserved to
                                procedures, and                                     the NRC because it addresses
                                drawings.                                           the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.113.................  Document control........  NRC.....................  This provision is reserved to
                                                                                    the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.115.................  Control of purchased      NRC.....................  This provision is reserved to
                                material, equipment,                                the NRC because it addresses
                                and services.                                       the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.

[[Page 21440]]

 
Sec.  71.117.................  Identification and        NRC.....................  This provision is reserved to
                                control of materials,                               the NRC because it addresses
                                parts, and components.                              the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.119.................  Control of special        NRC.....................  This provision is reserved to
                                processes.                                          the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.121.................  Internal Inspection.....  NRC.....................  This provision is reserved to
                                                                                    the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.123.................  Test control............  NRC.....................  This provision is reserved to
                                                                                    the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.125.................  Control of measuring and  NRC.....................  This provision is reserved to
                                test equipment.                                     the NRC because it addresses
                                                                                    the design, fabrication,
                                                                                    modification, and approval
                                                                                    of Type B packages.
Sec.  71.127.................  Handling, storage, and     D--for those States      For those States which have
                                shipping control.         which have no licensees   licensees that use Type B
                                                          that use Type B           packages, and have adopted
                                                          packages.                 the essential objectives of
                                                         [C]--for those States      Sec.  71.105(b), it is not
                                                          which have licensees      necessary for them to adopt
                                                          that use Type B           this provision again.
                                                          packages.
Sec.  71.129.................  Inspection, test, and     D.......................  This provision does not meet
                                operating status.                                   any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.131.................  Nonconforming materials,  D.......................  This provision does not meet
                                parts, or components.                               any of the criteria for
                                                                                    designations Category A, B,
                                                                                    C, or health and safety.
                                                                                    Thus, it does not need to be
                                                                                    adopted by Agreement States.
Sec.  71.133.................  Corrective action.......  D--for those States       This provision is designated
                                                          which have no licensees   Category C for those States
                                                          that use Type B           which have licensees that
                                                          packages.                 use Type B packages. This
                                                         C--for those States        provision is designated
                                                          which have licensees      Category C because the
                                                          that use Type B           quality assurance of Type B
                                                          packages.                 packages is an activity that
                                                                                    is needed in order to avoid
                                                                                    a nationwide gap in the
                                                                                    regulation of the
                                                                                    transportation of
                                                                                    radioactive materials. If
                                                                                    this provision is not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objective of
                                                                                    this provision is that each
                                                                                    licensee who uses a Type B
                                                                                    package shall establish
                                                                                    measures to assure that
                                                                                    conditions adverse to
                                                                                    quality, such as
                                                                                    deficiencies, deviations,
                                                                                    defective material and
                                                                                    equipment, and
                                                                                    nonconformances, are
                                                                                    promptly identified and
                                                                                    corrected.
Sec.  71.135.................  Quality assurance         D--for those States       This provision is designated
                                records.                  which have no licensees   a Category C for those
                                                          that use Type B           States which have licensees
                                                          packages.                 that use Type B packages.
                                                         C--for those States        This provision is designated
                                                          which have licensees      Category C because the
                                                          that use Type B           quality assurance of Type B
                                                          packages.                 packages is an activity that
                                                                                    is needed in order to avoid
                                                                                    a nationwide gap in the
                                                                                    regulation of the
                                                                                    transportation of
                                                                                    radioactive materials. If
                                                                                    this provision is not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objective of
                                                                                    this provision is that each
                                                                                    licensee who uses a Type B
                                                                                    package shall maintain
                                                                                    sufficient written records
                                                                                    to demonstrate compliance
                                                                                    with the quality assurance
                                                                                    program.
Sec.  71.137.................  Audits..................  D--for those States       This provision is designated
                                                          which have no licensees   a Category C for those
                                                          that use Type B           States which have licensees
                                                          packages.                 that use Type B packages.
                                                         C--for those States        This provision is designated
                                                          which have licensees      Category C because the
                                                          that use Type B           quality assurance of Type B
                                                          packages.                 packages is an activity that
                                                                                    is needed in order to avoid
                                                                                    a nationwide gap in the
                                                                                    regulation of the
                                                                                    transportation of
                                                                                    radioactive materials. If
                                                                                    this provision is not
                                                                                    adopted, this could result
                                                                                    in undesirable consequences
                                                                                    in multiple jurisdictions.
                                                                                    The essential objectives of
                                                                                    this provision are that each
                                                                                    licensee who uses a Type B
                                                                                    package shall carry out a
                                                                                    system of planned and
                                                                                    periodic audits to: (1)
                                                                                    verify compliance with all
                                                                                    aspects of the quality
                                                                                    assurance program, (2)
                                                                                    determine the effectiveness
                                                                                    of the program, and (3)
                                                                                    verify that the audits are
                                                                                    performed by appropriately
                                                                                    trained personnel.
Secs.  71.151 through 71.177.  Subpart I--Type B(DP)     NRC.....................  Subpart I is designated
                                Package Approval.                                   Category NRC because it
                                                                                    addresses Type B (DP)
                                                                                    package approval, an area
                                                                                    reserved to NRC's regulatory
                                                                                    authority.
Appendix A...................  Determination of A1 and   B.......................  This provision is designated
                                A2.                                                 a Category B because it
                                                                                    applies to activities that
                                                                                    have direct and significant
                                                                                    effects in multiple
                                                                                    jurisdictions.
----------------------------------------------------------------------------------------------------------------

VIII. Plain Language

    The Presidential Memorandum dated June 1, 1998, entitled ``Plain 
Language in Government Writing,'' directed that the Federal 
government's writing be in plain language. This memorandum was 
published June 10, 1998 (63 FR 31883). In complying with this 
directive, editorial changes have been made in the proposed revision to 
improve the organization and readability of the existing language of 
paragraphs being revised. These types of changes are not discussed 
further in this document. The NRC requests comments on the proposed 
rule specifically with respect to the clarity and effectiveness of the 
language used. Comments should be

[[Page 21441]]

sent to the address listed under the ADDRESSES heading.

IX. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995, 
Public Law 104-113, requires that Federal agencies use technical 
standards that are developed or adopted by voluntary consensus standard 
bodies unless the use of such a standard is inconsistent with 
applicable law or otherwise impractical. In this proposed rule, the NRC 
considered but decided not to adopt the ASME Code, Section III, 
Division 3, as described in Issue 14. However, the NRC is presenting 
amendments to its transportation regulations that would make them 
compatible with the IAEA transportation standards. This action does not 
constitute the establishment of a standard that establishes generally-
applicable requirements.

X. Environmental Assessment: Finding of No Significant 
Environmental Impact

    The Commission has prepared a draft environmental assessment 
entitled: Draft Environmental Assessment (EA) of Major Revision of 10 
CFR Part 71 (NUREG/CR-6711, March 2002), on this proposed regulation. 
The draft EA is available on the NRC rulemaking website and is also 
available for inspection in the NRC Public Document Room, 11555 
Rockville Pike, Room O-1F21, Rockville, MD. The Commission requests 
public comments on the draft EA. Comments on the draft EA may be 
submitted to the NRC as indicated under the ADDRESSES heading. The 
following is a brief summary of the draft EA.
    The EA grouped the proposed action into 19 different changes to 
Part 71, which could be adopted either all together as one list or 
independently in a partial list. Of these 19 changes, the following 
four meet the NRC's categorical exclusion criteria:
     Changes to Various Definitions (Issue 9);
     Expansion of Part 71 Quality Assurance Requirements to 
Certificate of Compliance (CoC) Holders (Issue 13);
     Change Authority for Dual-Purpose Package Certificate 
Holders (Issue 15); and
     Modifications of Event Reporting Requirements (Issue 19).
    None of the remaining 15 changes are expected to cause a 
significant impact to human health, safety, or the environment, whether 
promulgated altogether or individually. In fact, most of the changes 
would have negligible effects or result in slight improvements in 
health, safety, and environmental protection. In particular, the 
following changes are primarily administrative in nature, would not 
cause any new negative impacts, and would result in the beneficial 
effect of simplifying and/or harmonizing the NRC's regulations with TS-
R-1:
     Changing Part 71 to the International System of Units (SI) 
Only (Issue 1);
     Revision of A1 and A2 (Issue 3);
     A new requirement to display the Criticality Safety Index 
on shipping packages of fissile material (Issue 5);
     A provision to ``grandfather'' older shipping packages 
under the Part 71 requirements in existence when their Certificates of 
Compliance were issued (Issue 8); and
     Procedures for approval of special arrangements for 
shipment of special packages (Issue 12).
    The following changes would result in slight net improvements in 
health, safety, and environmental protection:
     Addition of uranium hexafluoride package requirements 
(Issue 4);
     Strengthening the requirements in Sec. 71.61 to ensure 
package containment in deep submersion scenarios (Issue 7);
     Adoption of the crush test for fissile material package 
design (Issue 10);
     Adoption of fissile material package design requirements 
for transport by aircraft (Issue 11); and
     Adoption of the ASME Code for spent fuel transportation 
casks (Issue 14).
    The proposal to change the existing 70 Bq/g (0.002 Ci/g) 
level to radionuclide-specific activity limits (Issue 2) is expected to 
have mixed, although overall minor, effects. For radionuclides with new 
exemption values that are lower than the current limit, there could be 
a decrease in the number of exempted shipments and a commensurate 
slight increase in the level of protection. For radionuclides with new 
exemption values that are higher than the current limit, there could be 
an increase in the number of exempted shipments and a commensurate 
slight increase in associated radiation exposures. However, IAEA and 
the NRC have determined that this change would not significantly 
increase the risk to individuals.
    The addition of the Type C package and low level dispersible 
material concepts (Issue 6) would result in mixed, although overall 
minor, effects. If the same number of packages are handled, the 
radiation doses to workers loading and unloading Type C packages 
shipped by air will be slightly higher than the doses to workers 
loading and unloading other kinds of packages shipped by other means. 
At the same time, ``incident-free'' doses during the shipping of Type C 
packages are expected to be slightly reduced compared to baseline 
conditions, while the risks associated with accidents during shipping 
could be slightly increased or decreased depending on the shipping 
scenario.
    Changes to transportation regulations for fissile materials 
actually consist of 17 individual recommendations for revisions to part 
71 (Issue 16). Ten of these recommendations are expected to result in 
no impact, as they simply clarify definitions, consolidate related 
requirements into single sections, or streamline the regulations. Four 
of the recommendations will result in small improvements to health, 
safety, and environmental protection by eliminating confusion among 
licensees and/or providing added assurance for critical safety. The 
last two recommendations, which would revise exemptions for low-level 
material and remove or modify provisions related to the shipment of Pu-
Be neutron sources, are expected to significantly improve criticality 
safety.
    Changes to the requirements for plutonium shipments in Sec. 71.63 
(PRM-71-12) could result in a slight increase in the probability and 
consequences of accidental releases, primarily when and if plutonium is 
shipped in liquid form. However, most plutonium shipments are either 
related to the disposition of plutonium wastes or to the production of 
mixed oxides, neither of which involve the shipment of a liquid 
solution of plutonium.
    No changes have been identified for the issue related to surface 
contamination limits as applied to spent fuel and high level waste 
(Issue 18). The issue was included in the proposed rule in response to 
Commission direction in SRM-SECY-00-0117. NRC is seeking input on 
whether the NRC should address this issue in future rulemaking 
activities. As a result, no regulatory options were developed, and 
therefore no environmental assessment conducted.
    The Commission has determined, under the National Environmental 
Policy Act of 1969, as amended, and the Commission's regulations in 
Subpart A of 10 CFR part 51, that this rule is not a major Federal 
action significantly affecting the quality of the human environment, 
and therefore an environmental impact statement (EIS) is not required.
    The Commission's ``Final Environmental Statement on the 
Transportation of Radioactive Material by Air and Other Modes,'' NUREG-

[[Page 21442]]

0170 \19\, dated December 1977, is NRC's generic EIS, covering all 
types of radioactive material transportation by all modes (road, rail, 
air, and water). From the Commission's latest survey of radioactive 
material shipments and their characteristics, ``Transport of 
Radioactive Material in the United States,'' SAND 84-7174, April 1985, 
the NRC concluded that current radioactive material shipments are not 
so different from those evaluated in NUREG-0170 as to invalidate the 
results or conclusions of that EIS. Environmental assessment of the 
impacts associated with this rulemaking is evaluated in ``Environmental 
Assessment of Major Revision to Packaging and Transportation of 
Radioactive Material Regulations (10 CFR part 71),'' dated February 
2000.
---------------------------------------------------------------------------

    \19\ Copies of NUREG-0170 may be purchased from the 
Superintendent of Documents, U.S. Government Printing Office, P.O. 
Box 37082, Washington, DC 20013-7082. Copies are also available from 
the National Technical Information Service, 5285 Port Royal Road, 
Springfield, VA 22161. A copy is also available for inspection and 
copying for a fee in the NRC Public Document Room, 11555 Rockville 
Pike, Room O-1F21, Rockville MD.
---------------------------------------------------------------------------

    NUREG-0170 established the nonaccident related radiation exposures 
associated with transportation of radioactive material in the United 
States as 98 person-Sv (9800 person-rem) which, based on the 
conservative linear radiation dose hypothesis, resulted in a maximum of 
1.7 genetic effects and 1.2 latent cancer effects per year. More than 
half this impact resulted from shipment of medical-use radioactive 
materials. Accident related impacts were established at a maximum of 
one genetic effect and one latent cancer fatality for 200 years of 
transporting radioactive materials. The principal nonradiological 
impacts were found to be two injuries per year, and less than one 
accidental death per 4 years. In contrast, nonaccident related 
radiation exposures and accident related impacts associated with this 
rulemaking would not change from the impact of the current Part 71 
requirements (i.e., no increase or decrease). Nonradiological traffic 
injuries and nonradiological traffic deaths would not change. These 
impacts are judged to be insignificant compared with the baseline 
impacts established in NUREG-0170.
    The environmental assessment and finding of no significant impact 
on which this determination is based are available, for inspection, at 
the NRC Public Document Room, 11555 Rockville Pike, Room O-1F21, 
Rockville, MD. The environmental assessment is also available on the 
NRC rulemaking website.

XI. Paperwork Reduction Act Statement

    The proposed rule would amend information collection requirements 
that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 
et seq.). This rule has been submitted to the Office of Management and 
Budget for review and approval of the information collection 
requirements.
    The burden to the public for these information collections is 
estimated to average 16.3 hours per response, including the time for 
reviewing instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
information collection. The U.S. Nuclear Regulatory Commission is 
seeking public comment on the potential impact of the information 
collections contained in the proposed rule and on the following issues:
    1. Is the proposed information collection necessary for the proper 
performance of the functions of the NRC, including whether the 
information will have practical utility?
    2. Is the estimate of burden accurate?
    3. Is there a way to enhance the quality, utility, and clarity of 
the information to be collected?
    4. How can the burden of the information collection be minimized, 
including the use of automated collection techniques?
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden, to the 
Records Management Branch (T-6E6), U.S. Nuclear Regulatory Commission, 
Washington DC 20555-0001, or by Internet electronic mail at 
[email protected]; and to the Desk Officer, Office of Information 
and Regulatory Affairs, NEOB-10202 (3150-0008), Office of Management 
and Budget, Washington, DC 20503.
    Comments to OMB on the information collections or on the above 
issues should be submitted by May 30, 2002. Comments received after 
this date will be considered if it is practical to do so, but assurance 
of consideration cannot be given to comments received after this date.

Public Protection Notification

    If a means used to impose an information collection does not 
display a currently valid OMB control number, the NRC may not conduct 
or sponsor, and person is not required to respond to, the information 
collection.

XII. Regulatory Analysis

    The Commission has prepared a draft regulatory analysis entitled 
``Draft Regulatory Analysis of Major Revision of 10 CFR part 71--
Proposed Rule, NUREG/CR-6713, March 2002.'' To support the discussions 
of the proposed changes, selected material from this regulatory 
analysis has been included earlier under each issue. The analysis 
examines the costs and benefits of the alternatives considered by the 
Commission. The draft regulatory analysis is available on the NRC 
rulemaking website, also available for inspection at the NRC Public 
Document Room, 11555 Rockville Pike, Room O-1F21, Rockville, MD. The 
Commission requests public comments on the draft regulatory analysis. 
Comments on the draft analysis may be submitted to the NRC as indicated 
under the ADDRESSES heading.

XIII. Regulatory Flexibility Act Certification

    In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 
605(b)), the Commission certifies that this rule will not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This proposed rule affects NRC licensees, including 
operators of nuclear power plants, who transport or deliver to a 
carrier for transport, relatively large quantities of radioactive 
material in a single package. These companies do not generally fall 
within the scope of the definition of ``small entities'' set forth in 
the Regulatory Flexibility Act or the size standards adopted by the NRC 
(10 CFR 2.810).

XIV. Backfit Analysis

    The NRC has determined that the backfit rule does not apply to this 
proposed rule; therefore, a backfit analysis is not required for this 
proposed rule because these amendments do not involve any provisions 
that would require backfits as defined in 10 CFR Chapter I.

List of Subjects in 10 CFR Part 71

    Criminal penalties, Hazardous materials transportation, Nuclear 
materials, Packaging and containers, Reporting and recordkeeping 
requirements.

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended; the Energy Reorganization 
Act of 1974, as amended; and 5 U.S.C. 553, the Commission is proposing 
to revise 10 CFR Part 71 as follows:

[[Page 21443]]

PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

    1. The authority citation for Part 71 continues to read as follows:

    Authority:  Secs. 53, 57, 62, 63, 81, 161, 182, 183, 68 Stat. 
930, 932, 933, 935, 948, 953, 954, as amended, sec. 1701, 106 Stat. 
2951, 2952, 2953 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 
2232, 2233, 2297f); secs. 201, as amended, 202, 206, 88 Stat. 1242, 
as amended, 1244, 1246 (42 U.S.C. 5841, 5842, 5846);

    Section 71.97 also issued under sec. 301, Pub. L. 96-295, 94 
Stat. 789-790.

    2. Subparts A, B, and C to Part 71 are revised to read as follows:
Subpart A--General Provisions
Sec.
71.0   Purpose and scope.
71.1   Communications and records.
71.2   Interpretations.
71.3   Requirement for license.
71.4   Definitions.
71.5   Transportation of licensed material.
71.6   Information collection requirements: OMB approval.
71.7   Completeness and accuracy of information.
71.8   Deliberate misconduct.
71.9   Employee protection.
71.10   Public inspection of application.
71.11   [Reserved]
Subpart B--Exemptions
71.12   Specific exemptions.
71.13   Exemption of physicians.
71.14   Exemption for low-level materials.
71.15   Exemption from classification as fissile material.
71.16   [Reserved.]
Subpart C--General Licenses
71.17   General license: NRC-approved package.
71.18   General license: NRC-approved Type B(DP) package.
71.19   Previously approved package.
71.20   General license: DOT specification container.
71.21   General License: Use of foreign approved package.
71.22   General license: Fissile material.
71.23   General license: Plutonium-beryllium special form material.
71.24   [Reserved]
71.25   [Reserved]

Subpart A--General Provisions


Sec. 71.0  Purpose and scope.

    (a) This part establishes --
    (1) Requirements for packaging, preparation for shipment, and 
transportation of licensed material; and
    (2) Procedures and standards for NRC approval of packaging and 
shipping procedures for fissile material and for a quantity of other 
licensed material in excess of a Type A quantity.
    (b) The packaging and transport of licensed material are also 
subject to other parts of this chapter (e.g., 10 CFR parts 20, 21, 30, 
40, 70, and 73) and to the regulations of other agencies (e.g., the 
U.S. Department of Transportation (DOT) and the U.S. Postal Service) 
\1\ having jurisdiction over means of transport. The requirements of 
this part are in addition to, and not in substitution for, other 
requirements.
---------------------------------------------------------------------------

    \1\ Department of Transportation regulations in 49 CFR chapter 
I; Postal Service manual (Domestic Manual), Section 124, which is 
incorporated by reference at 39 CFR 111.1.
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    (c) The regulations in this part apply to any licensee authorized 
by specific or general license issued by the Commission to receive, 
possess, use, or transfer licensed material, if the licensee delivers 
that material to a carrier for transport, transports the material 
outside the site of usage as specified in the NRC license, or 
transports that material on public highways. No provision of this part 
authorizes possession of licensed material.
    (d)(1) Exemptions from the requirement for license in Sec. 71.3 are 
specified in Sec. 71.14. General licenses for which no NRC package 
approval is required are issued in Secs. 71.20 through 71.23. The 
general license in Sec. 71.17 requires that an NRC certificate of 
compliance or other package approval be issued for the package to be 
used under this general license. The general license in Sec. 71.18 
requires that an NRC certificate of compliance or other package 
approval be issued for the Type B(DP) package to be used under this 
general license.
    (2) Application for package approval, other than Type B(DP) 
packages, must be completed in accordance with subpart D of this part, 
demonstrating that the design of the package to be used satisfies the 
package approval standards contained in subpart E of this part, as 
related to the tests of subpart F of this part.
    (3) Application for Type B(DP) package approval must be completed 
in accordance with subpart I of this part, demonstrating that the 
design of the package to be used satisfies the applicable package 
approval standards contained in subpart E of this part, as related to 
the tests of subpart F of this part.
    (4) A licensee transporting licensed material, or delivering 
licensed material to a carrier for transport, shall comply with the 
operating control requirements of subpart G of this part; the quality 
assurance requirements of subpart H of this part; and the general 
provisions of subpart A of this part, including DOT regulations 
referenced in Sec. 71.5.
    (e) The regulations of this part apply to any person holding or 
applying for a certificate of compliance, issued pursuant to this part, 
for a package intended for the transportation of radioactive material, 
outside the confines of a licensee's facility or authorized place of 
use.
    (f) The regulations in this part apply to any person required to 
obtain a certificate of compliance, or an approved compliance plan, 
pursuant to part 76 of this chapter, if the person delivers radioactive 
material to a common or contract carrier for transport or transports 
the material outside the confines of the person's plant or other 
authorized place of use.
    (g) This part also gives notice to all persons who knowingly 
provide to any licensee, certificate holder, quality assurance program 
approval holder, applicant for a license, certificate, or quality 
assurance program approval, or to a contractor, or subcontractor of any 
of them, components, equipment, materials, or other goods or services, 
that relate to a licensee's, certificate holder's, quality assurance 
program approval holder's, or applicant's activities subject to this 
part, that they may be individually subject to NRC enforcement action 
for violation of Sec. 71.8.


Sec. 71.1  Communications and records.

    (a) Except where otherwise specified, all communications and 
reports concerning the regulations in this part and applications filed 
under them should be addressed to the U.S. Nuclear Regulatory 
Commission, ATTN: Document Control Desk, Washington, DC 20555-0001. 
Written communications, reports, and applications may be delivered in 
person to the U.S. NRC, ATTN: Document Control Desk, at One White Flint 
North, 11555 Rockville Pike, Rockville, MD 20852-2738 between 7:30 a.m. 
and 4:15 p.m., Federal workdays. If the submittal deadline date falls 
on a Saturday, Sunday, or a Federal holiday, the next Federal workday 
becomes the official due date.
    (b) Each record required by this part must be legible throughout 
the retention period specified by each Commission regulation. The 
record may be the original or a reproduced copy or a microform provided 
that the copy or microform is authenticated by authorized personnel and 
that the microform is capable of producing a clear copy throughout the 
required retention period. The record may also be stored in electronic 
media with the capability for producing legible, accurate, and complete 
records during the required retention period. Records such as letters, 
drawings, and

[[Page 21444]]

specifications must include all pertinent information such as stamps, 
initials, and signatures. The licensee shall maintain adequate 
safeguards against tampering with and loss of records.


Sec. 71.2  Interpretations.

    Except as specifically authorized by the Commission in writing, no 
interpretation of the meaning of the regulations in this part by any 
officer or employee of the Commission, other than a written 
interpretation by the General Counsel, will be recognized to be binding 
upon the Commission.


Sec. 71.3  Requirement for license.

    Except as authorized in a general license or a specific license 
issued by the Commission, or as exempted in this part, no licensee 
may--
    (a) Deliver licensed material to a carrier for transport; or
    (b) Transport licensed material.


Sec. 71.4  Definitions.

    The following terms are as defined here for the purpose of this 
part. To ensure compatibility with international transportation 
standards, all limits in this part are given in terms of dual units: 
The International System of Units (SI) followed or preceded by U.S. 
standard or customary units. The U.S. customary units are not exact 
equivalents but are rounded to a convenient value, providing a 
functionally equivalent unit. For the purpose of this part, either unit 
may be used.
    A1 means the maximum activity of special form 
radioactive material permitted in a Type A package. This value is 
either listed in Appendix A, Table A-1, of this part, or may be derived 
in accordance with the procedures prescribed in Appendix A of this 
part.
    A2 means the maximum activity of radioactive material, 
other than special form material, LSA, and SCO material, permitted in a 
Type A package. This value is either listed in Appendix A, Table A-1, 
of this part, or may be derived in accordance with the procedures 
prescribed in Appendix A of this part.
    Carrier means a person engaged in the transportation of passengers 
or property by land or water as a common, contract, or private carrier, 
or by civil aircraft.
    Certificate holder means a person who has been issued a certificate 
of compliance or other package approval by the Commission.
    Certificate of Compliance (CoC) means the certificate issued by the 
Commission under either subpart D or I of this part which approves the 
design of a package for the transportation of radioactive material.
    Close reflection by water means immediate contact by water of 
sufficient thickness for maximum reflection of neutrons.
    Containment system means the assembly of components of the 
packaging intended to retain the radioactive material during transport.
    Conveyance means:
    (1) For transport by public highway or rail any transport vehicle 
or large freight container;
    (2) For transport by water any vessel, or any hold, compartment, or 
defined deck area of a vessel including any transport vehicle on board 
the vessel; and
    (3) For transport by aircraft any aircraft.
    Criticality Safety Index (CSI) means the dimensionless number 
(rounded up to the next tenth) assigned to and placed on the label of a 
fissile material package, to designate the degree of control of 
accumulation of packages containing fissile material during 
transportation. Determination of the criticality safety index is 
described in Secs. 71.22, 71.23, and 71.59.
    Deuterium means, for the purposes of Secs. 71.15 and 71.22, the 
definition of Deuterium as found in Sec. 110.2 of this chapter.
    DOT means the U.S. Department of Transportation.
    Exclusive use means the sole use by a single consignor of a 
conveyance for which all initial, intermediate, and final loading and 
unloading are carried out in accordance with the direction of the 
consignor or consignee. The consignor and the carrier must ensure that 
any loading or unloading is performed by personnel having radiological 
training and resources appropriate for safe handling of the 
consignment. The consignor must issue specific instructions, in 
writing, for maintenance of exclusive use shipment controls, and 
include them with the shipping paper information provided to the 
carrier by the consignor.
    Fissile material means the radionuclides uranium-233, uranium-235, 
plutonium-239, and plutonium-241, or any combination of these 
radionuclides. Fissile material means the fissile nuclides themselves, 
not material containing fissile nuclides. Unirradiated natural uranium 
and depleted uranium and natural uranium or depleted uranium, that has 
been irradiated in thermal reactors only, are not included in this 
definition. Certain exclusions from fissile material controls are 
provided in Sec. 71.15.
    Graphite means, for the purposes of Secs. 71.15 and 71.22, the 
definition of Nuclear grade graphite as found in Sec. 110.2 of this 
chapter.
    Licensed material means by-product, source, or special nuclear 
material received, possessed, used, or transferred under a general or 
specific license issued by the Commission pursuant to the regulations 
in this chapter.
    Low Specific Activity (LSA) material means radioactive material 
with limited specific activity that satisfies the descriptions and 
limits set forth in this definition. Shielding materials surrounding 
the LSA material may not be considered in determining the estimated 
average specific activity of the package contents. LSA material must be 
in one of three groups:
    (1) LSA--I.
    (i) Ores containing only naturally occurring radionuclides (e.g., 
uranium, thorium) and uranium or thorium concentrates of such ores;
    (ii) Solid unirradiated natural uranium or depleted uranium or 
natural thorium or their solid or liquid compounds or mixtures;
    (iii) Radioactive material, other than fissile material, for which 
the A2 value is unlimited; or
    (iv) Mill tailings, contaminated earth, concrete, rubble, other 
debris, and activated material in which the radioactive material is 
essentially uniformly distributed, and the average specific activity 
does not exceed 10-6 A2/g.
    (2) LSA--II.
    (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
    (ii) Material in which the radioactive material is distributed 
throughout, and the average specific activity does not exceed 
10-4 A2/g for solids and gases, and 10-5 
A2/g for liquids.
    (3) LSA--III. Solids (e.g., consolidated wastes, activated 
materials) that satisfy the requirements of Sec. 71.77, in which:
    (i) The radioactive material is distributed throughout a solid or a 
collection of solid objects, or is essentially uniformly distributed in 
a solid compact binding agent (such as concrete, bitumen, ceramic, 
etc.);
    (ii) The radioactive material is relatively insoluble, or it is 
intrinsically contained in a relatively insoluble material, so that, 
even under loss of packaging, the loss of radioactive material per 
package by leaching, when placed in water for 7 days, would not exceed 
0.1 A2; and
    (iii) The average specific activity of the solid does not exceed 2 
x 10-3 A2/g.
    Low toxicity alpha emitters means natural uranium, depleted 
uranium,

[[Page 21445]]

natural thorium; uranium-235, uranium-238, thorium-232, thorium-228 or 
thorium-230 when contained in ores or physical or chemical concentrates 
or tailings; or alpha emitters with a half-life of less than 10 days.
    Maximum normal operating pressure means the maximum gauge pressure 
that would develop in the containment system in a period of 1 year 
under the heat condition specified in Sec. 71.71(c)(1), in the absence 
of venting, external cooling by an ancillary system, or operational 
controls during transport.
    Natural thorium means thorium with the naturally occurring 
distribution of thorium isotopes (essentially 100 weight percent 
thorium-232).
    Normal form radioactive material means radioactive material that 
has not been demonstrated to qualify as ``special form radioactive 
material.''
    Optimum interspersed hydrogenous moderation means the presence of 
hydrogenous material between packages to such an extent that the 
maximum nuclear reactivity results.
    Package means the packaging together with its radioactive contents 
as presented for transport.
    (1) Fissile material package or Type AF package, Type BF package, 
Type B(U)F package, or Type B(M)F package means a fissile material 
packaging together with its fissile material contents.
    (2) Type A package means a Type A packaging together with its 
radioactive contents. A Type A package is defined and must comply with 
the DOT regulations in 49 CFR part 173.
    (3) Type B package means a Type B packaging together with its 
radioactive contents. On approval, a Type B package design is 
designated by NRC as B(U) unless the package has a maximum normal 
operating pressure of more than 700 kPa (100 lbs/in\2\) gauge or a 
pressure relief device that would allow the release of radioactive 
material to the environment under the tests specified in Sec. 71.73 
(hypothetical accident conditions), in which case it will receive a 
designation B(M). B(U) refers to the need for unilateral approval of 
international shipments; B(M) refers to the need for multilateral 
approval of international shipments. There is no distinction made in 
how packages with these designations may be used in domestic 
transportation. To determine their distinction for international 
transportation, see DOT regulations in 49 CFR part 173. A Type B 
package approved before September 6, 1983, was designated only as Type 
B. Limitations on its use are specified in Sec. 71.19.
    (4) Type B(DP) package means a Type B(DP) packaging together with 
its radioactive contents. A Type B(DP) package is a dual-purpose 
package intended for both the transportation and storage of spent fuel. 
A Type B(DP) package is also a fissile material package. A Type B(DP) 
package is issued both a certificate of compliance approving the design 
of a spent-fuel transportation package, in accordance with subpart I of 
this part, and a certificate of compliance approving the design of a 
spent fuel storage cask, in accordance with subpart L of part 72 of 
this chapter.
    Packaging means the assembly of components necessary to ensure 
compliance with the packaging requirements of this part. It may consist 
of one or more receptacles, absorbent materials, spacing structures, 
thermal insulation, radiation shielding, and devices for cooling or 
absorbing mechanical shocks. The vehicle, tie-down system, and 
auxiliary equipment may be designated as part of the packaging.
    Special form radioactive material means radioactive material that 
satisfies the following conditions:
    (1) It is either a single solid piece or is contained in a sealed 
capsule that can be opened only by destroying the capsule;
    (2) The piece or capsule has at least one dimension not less than 5 
mm (0.2 in); and
    (3) It satisfies the requirements of Sec. 71.75. A special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on June 30, 1983 (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before July 1, 1985, and a special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on March 31, 1996 (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before April 1, 1998, may continue to be 
used. Any other special form encapsulation must meet the specifications 
of this definition.
    Specific activity of a radionuclide means the radioactivity of the 
radionuclide per unit mass of that nuclide. The specific activity of a 
material in which the radionuclide is essentially uniformly distributed 
is the radioactivity per unit mass of the material.
    Spent nuclear fuel or Spent fuel means fuel that has been withdrawn 
from a nuclear reactor following irradiation, has undergone at least 
one year's decay since being used as a source of energy in a power 
reactor, and has not been chemically separated into its constituent 
elements by reprocessing. Spent fuel includes the special nuclear 
material, byproduct material, source material, and other radioactive 
materials associated with fuel assemblies.
    State means a State of the United States, the District of Columbia, 
the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American 
Samoa, and the Commonwealth of the Northern Mariana Islands.
    Structures, systems, and components important to safety (SSCs) 
means those features of a Type B(DP) package whose functions are--
    (1) To maintain the conditions required to safely transport the 
package's contents;
    (2) To prevent damage to the package during transport; or
    (3) To provide reasonable assurance that the radioactive material 
contents can be received, handled, transported, and retrieved without 
undue risk to public health and safety and the environment.
    Surface Contaminated Object (SCO) means a solid object that is not 
itself classed as radioactive material, but which has radioactive 
material distributed on any of its surfaces. SCO must be in one of two 
groups with surface activity not exceeding the following limits:
    (1) SCO--I: A solid object on which:
    (i) The nonfixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 4 Bq/cm\2\ (10-4 microcurie/cm\2\) for beta and 
gamma and low toxicity alpha emitters, or 0.4 Bq/cm\2\ (10-5 
microcurie/cm\2\) for all other alpha emitters;
    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 4 x 10\4\ Bq/cm\2\ (1.0 microcurie/cm\2\) for beta and gamma 
and low toxicity alpha emitters, or 4 x 10\3\ Bq/cm\2\ (0.1 microcurie/
cm\2\) for all other alpha emitters; and
    (iii) The nonfixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm\2\ (or the area of the 
surface if less than 300 cm\2\) does not exceed 4 x 10\4\ Bq/cm\2\ (1 
microcurie/cm\2\) for beta and gamma and low toxicity alpha emitters, 
or 4 x 10\3\ Bq/cm\2\ (0.1 microcurie/cm\2\) for all other alpha 
emitters.
    (2) SCO--II: A solid object on which the limits for SCO--I are 
exceeded and on which:
    (i) The nonfixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 400 Bq/cm\2\ (10-2 microcurie/cm\2\) for beta and 
gamma and low toxicity alpha emitters or 40 Bq/cm\2\ (10-3 
microcurie/cm\2\) for all other alpha emitters;

[[Page 21446]]

    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 8 x 10\5\ Bq/cm\2\ (20 microcuries/cm\2\) for beta and gamma 
and low toxicity alpha emitters, or 8 x 10\4\ Bq/cm\2\ (2 microcuries/
cm\2\) for all other alpha emitters; and
    (iii) The nonfixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm\2\ (or the area of the 
surface if less than 300 cm\2\) does not exceed 8 x 10\5\ Bq/cm\2\ (20 
microcuries/cm\2\) for beta and gamma and low toxicity alpha emitters, 
or 8 x 10\4\ Bq/cm\2\ (2 microcuries/cm\2\) for all other alpha 
emitters.
    Transport index (TI) means the dimensionless number (rounded up to 
the next tenth) placed on the label of a package, to designate the 
degree of control to be exercised by the carrier during transportation. 
The transport index is the number determined by multiplying the maximum 
radiation level in millisievert (mSv) per hour at 1 meter (3.3 ft) from 
the external surface of the package by 100 (equivalent to the maximum 
radiation level in millirem per hour at 1 meter (3.3 ft)).
    Type A quantity means a quantity of radioactive material, the 
aggregate radioactivity of which does not exceed A1 for 
special form radioactive material, or A2, for normal form 
radioactive material, where A1 and A\2\ are given in Table 
A--1 of this part, or may be determined by procedures described in 
Appendix A of this part.
    Type B quantity means a quantity of radioactive material greater 
than a Type A quantity.
    Uranium--natural, depleted, enriched:
    (1) Natural uranium means uranium with the naturally occurring 
distribution of uranium isotopes (approximately 0.711 weight percent 
uranium-235, and the remainder by weight essentially uranium-238).
    (2) Depleted uranium means uranium containing less uranium-235 than 
the naturally occurring distribution of uranium isotopes.
    (3) Enriched uranium means uranium containing more uranium-235 than 
the naturally occurring distribution of uranium isotopes.


Sec. 71.5  Transportation of licensed material.

    (a) Each licensee who transports licensed material outside the site 
of usage, as specified in the NRC license, or where transport is on 
public highways, or who delivers licensed material to a carrier for 
transport, shall comply with the applicable requirements of the DOT 
regulations in 49 CFR parts 170 through 189 appropriate to the mode of 
transport.
    (1) The licensee shall particularly note DOT regulations in the 
following areas:
    (i) Packaging--49 CFR part 173: subparts A and B and I.
    (ii) Marking and labeling--49 CFR part 172: subpart D, 
Secs. 172.400 through 172.407, Secs. 172.436 through 172.440, and 
subpart E.
    (iii) Placarding--49 CFR part 172: subpart F, especially 
Secs. 172.500 through 172.519, 172.556, and appendices B and C.
    (iv) Accident reporting--49 CFR part 171: Secs. 171.15 and 171.16.
    (v) Shipping papers and emergency information--49 CFR part 172: 
subparts C and G.
    (vi) Hazardous material employee training--49 CFR part 172: subpart 
H.
    (vii) Hazardous material shipper/carrier registration--49 CFR part 
107: subpart G.
    (2) The licensee shall also note DOT regulations pertaining to the 
following modes of transportation:
    (i) Rail--49 CFR part 174: subparts A through D and K.
    (ii) Air--49 CFR part 175.
    (iii) Vessel--49 CFR part 176: subparts A through F and M.
    (iv) Public Highway--49 CFR part 177 and parts 390 through 397.
    (b) If DOT regulations are not applicable to a shipment of licensed 
material, the licensee shall conform to the standards and requirements 
of the DOT specified in paragraph (a) of this section to the same 
extent as if the shipment or transportation were subject to DOT 
regulations. A request for modification, waiver, or exemption from 
those requirements, and any notification referred to in those 
requirements, must be filed with, or made to, the Director, Office of 
Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001.


Sec. 71.6  Information collection requirements: OMB approval.

    (a) The Nuclear Regulatory Commission has submitted the information 
collection requirements contained in this part to the Office of 
Management and Budget (OMB) for approval as required by the Paperwork 
Reduction Act (44 U.S.C. 3501 et seq.). The NRC may not conduct or 
sponsor, and a person is not required to respond to, a collection of 
information unless it displays a currently valid OMB control number. 
OMB has approved the information collection requirements contained in 
this part under control number 3150-0008.
    (b) The approved information collection requirements contained in 
this part appear in Secs. 71.5, 71.7, 71.9, 71.12, 71.17, 71.18, 71.19, 
71.20, 71.31, 71.33, 71.35, 71.37, 71.38, 71.39, 71.41, 71.47, 71.85, 
71.87, 71.89, 71.91, 71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 
71.107, 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 
71.125, 71.127, 71.129, 71.131, 71.133, 71.135, 71.137, 71.151, 71.153, 
71.155, 71.157, 71.159, 71.161, 71.165, 71.167, 71.171, 71.173, 71.175, 
71.177, and Appendix A.


Sec. 71.7  Completeness and accuracy of information.

    (a) Information provided to the Commission by a licensee, 
certificate holder, or an applicant for a license or CoC; or 
information required by statute or by the Commission's regulations, 
orders, license or CoC conditions, to be maintained by the licensee or 
certificate holder, must be complete and accurate in all material 
respects.
    (b) Each licensee, certificate holder, or applicant for a license 
or CoC must notify the Commission of information identified by the 
licensee, certificate holder, or applicant for a license or CoC as 
having, for the regulated activity, a significant implication for 
public health and safety or common defense and security. A licensee, 
certificate holder, or an applicant for a license or CoC violates this 
paragraph only if the licensee, certificate holder, or applicant for a 
license or CoC fails to notify the Commission of information that the 
licensee, certificate holder, or applicant for a license or CoC has 
identified as having a significant implication for public health and 
safety or common defense and security. Notification must be provided to 
the Administrator of the appropriate Regional Office within two working 
days of identifying the information. This requirement is not applicable 
to information which is already required to be provided to the 
Commission by other reporting or updating requirements.


Sec. 71.8  Deliberate misconduct.

    (a) This section applies to any--
    (1) Licensee;
    (2) Certificate holder;
    (3) Quality assurance program approval holder;
    (4) Applicant for a license, certificate, or quality assurance 
program approval;
    (5) Contractor (including a supplier or consultant) or 
subcontractor, to any person identified in paragraphs (a)(4) of this 
section; or
    (6) Employees of any person identified in paragraphs (a)(1) through 
(a)(5) of this section.

[[Page 21447]]

    (b) A person identified in paragraph (a) of this section who 
knowingly provides to any entity, listed in paragraphs (a)(1) through 
(a)(5) of this section, any components, materials, or other goods or 
services that relate to a licensee's, certificate holder's, quality 
assurance program approval holder's or applicant's activities subject 
to this part may not:
    (1) Engage in deliberate misconduct that causes or would have 
caused, if not detected, a licensee, certificate holder, quality 
assurance program approval holder, or any applicant to be in violation 
of any rule, regulation, or order; or any term, condition or limitation 
of any license, certificate or approval issued by the Commission; or
    (2) Deliberately submit to the NRC, a licensee, a certificate 
holder, quality assurance program approval holder, an applicant for a 
license, certificate or quality assurance program approval, or a 
licensee's, applicant's, certificate holder's, or quality assurance 
program approval holder's contractor or subcontractor, information that 
the person submitting the information knows to be incomplete or 
inaccurate in some respect material to the NRC.
    (c) A person who violates paragraph (b)(1) or (b)(2) of this 
section may be subject to enforcement action in accordance with the 
procedures in 10 CFR part 2, subpart B.
    (d) For the purposes of paragraph (b)(1) of this section, 
deliberate misconduct by a person means an intentional act or omission 
that the person knows:
    (1) Would cause a licensee, certificate holder, quality assurance 
program approval holder, or applicant for a license, certificate, or 
quality assurance program approval to be in violation of any rule, 
regulation, or order; or any term, condition, or limitation of any 
license or certificate issued by the Commission; or
    (2) Constitutes a violation of a requirement, procedure, 
instruction, contract, purchase order, or policy of a licensee, 
certificate holder, quality assurance program approval holder, 
applicant, or the contractor or subcontractor of any of them.


Sec. 71.9  Employee protection.

    (a) Discrimination by a Commission licensee, certificate holder, an 
applicant for a Commission license or a CoC, or a contractor or 
subcontractor of any of these, against an employee for engaging in 
certain protected activities, is prohibited. Discrimination includes 
discharge and other actions that relate to compensation, terms, 
conditions, or privileges of employment. The protected activities are 
established in section 211 of the Energy Reorganization Act of 1974, as 
amended, and in general are related to the administration or 
enforcement of a requirement imposed under the Atomic Energy Act of 
1954, as amended, or the Energy Reorganization Act of 1974, as amended.
    (1) The protected activities include, but are not limited to:
    (i) Providing the Commission or his or her employer information 
about alleged violations of either of the statutes named in paragraph 
(a) of this section or possible violations of requirements imposed 
under either of those statutes;
    (ii) Refusing to engage in any practice made unlawful under either 
of the statutes named in paragraph (a) of this section or under the 
requirements in 10 CFR chapter I if the employee has identified the 
alleged illegality to the employer;
    (iii) Requesting the Commission to institute action against his or 
her employer for the administration or enforcement of the requirements 
in 10 CFR chapter I.
    (iv) Testifying in any Commission proceeding, or before Congress, 
or at any Federal or State proceeding regarding any provision (or 
proposed provision) of either of the statutes named in paragraph (a) of 
this section; and
    (v) Assisting or participating in, or is about to assist or 
participate in, these activities.
    (2) These activities are protected even if no formal proceeding is 
actually initiated as a result of the employee's assistance or 
participation.
    (3) This section has no application to any employee alleging 
discrimination prohibited by this section who, acting without direction 
from his or her employer (or the employer's agent), deliberately causes 
a violation of any requirement of the Energy Reorganization Act of 
1974, as amended, or the Atomic Energy Act of 1954, as amended.
    (b) Any employee who believes that he or she has been discharged or 
otherwise discriminated against by any person for engaging in protected 
activities specified in paragraph (a)(1) of this section may seek a 
remedy for the discharge or discrimination through an administrative 
proceeding in the Department of Labor. The administrative proceeding 
must be initiated within 180 days after an alleged violation occurs. 
The employee may do this by filing a complaint alleging the violation 
with the Department of Labor, Employment Standards Administration, Wage 
and Hour Division. The Department of Labor may order reinstatement, 
back pay, and compensatory damages.
    (c) A violation of paragraph (a), (e), or (f) of this section by a 
Commission licensee, certificate holder, applicant for a Commission 
license or a CoC, or a contractor or subcontractor of any of these may 
be grounds for:
    (1) Denial, revocation, or suspension of the license or the CoC;
    (2) Imposition of a civil penalty on the licensee or applicant; or
    (3) Other enforcement action.
    (d) Actions taken by an employer, or others, which adversely affect 
an employee may be predicated upon nondiscriminatory grounds. The 
prohibition applies when the adverse action occurs because the employee 
has engaged in protected activities. An employee's engagement in 
protected activities does not automatically render him or her immune 
from discharge or discipline for legitimate reasons or from adverse 
action dictated by nonprohibited considerations.
    (e)(1) Each licensee, certificate holder, and applicant for a 
license or CoC must prominently post the current revision of NRC Form 
3, ``Notice to Employees,'' referenced in Sec. 19.11(c) of this 
chapter. This form must be posted at locations sufficient to permit 
employees protected by this section to observe a copy on the way to or 
from their place of work. The premises must be posted not later than 30 
days after an application is docketed and remain posted while the 
application is pending before the Commission, during the term of the 
license or CoC, and for 30 days following license or CoC termination.
    (2) Copies of NRC Form 3 may be obtained by writing to the Regional 
Administrator of the appropriate U.S. Nuclear Regulatory Commission 
Regional Office listed in Appendix D to part 20 of this chapter or by 
calling the NRC Publishing Services Branch at 301-415-5877.
    (f) No agreement affecting the compensation, terms, conditions, or 
privileges of employment, including an agreement to settle a complaint 
filed by an employee with the Department of Labor pursuant to section 
211 of the Energy Reorganization Act of 1974, as amended, may contain 
any provision which would prohibit, restrict, or otherwise discourage 
an employee from participating in a protected activity as defined in 
paragraph (a)(1) of this section including, but not limited to, 
providing information to the NRC or to his or her employer on potential 
violations or other matters within NRC's regulatory responsibilities.

[[Page 21448]]

Sec. 71.10  Public inspection of application.

    Applications for approval of a package design under this part, 
which are submitted to the Commission, may be made available for public 
inspection, in accordance with provisions of parts 2 and 9 of this 
chapter. This includes an application to amend or revise an existing 
package design, any associated documents and drawings submitted with 
the application, and any responses to NRC requests for additional 
information.


Sec. 71.11  [Reserved]

Subpart B--Exemptions


Sec. 71.12  Specific exemptions.

    On application of any interested person or on its own initiative, 
the Commission may grant any exemption from the requirements of the 
regulations in this part that it determines is authorized by law and 
will not endanger life or property nor the common defense and security.


Sec. 71.13  Exemption of physicians.

    Any physician licensed by a State to dispense drugs in the practice 
of medicine is exempt from Sec. 71.5 with respect to transport by the 
physician of licensed material for use in the practice of medicine. 
However, any physician operating under this exemption must be licensed 
under 10 CFR part 35 or the equivalent Agreement State regulations.


Sec. 71.14  Exemption for low-level materials.

    (a) A licensee is exempt from all the requirements of this part 
with respect to shipment or carriage of the following low-level 
materials:
    (1) Natural material and ores containing naturally occurring 
radionuclides that are not intended to be processed for use of these 
radionuclides, provided the activity concentration of the material does 
not exceed 10 times the values specified in Appendix A of this part.
    (2) Materials for which the activity concentration is not greater 
than the activity concentration values specified in Appendix A of this 
part, or for which the consignment activity is not greater than the 
limit for an exempt consignment found in Appendix A of this part.
    (b) A licensee is exempt from all the requirements of this part, 
other than Secs. 71.5 and 71.88, with respect to shipment or carriage 
of the following packages, provided the packages do not contain any 
fissile material, or the material is exempt from classification as 
fissile material under Sec. 71.15:
    (1) The package contains no more than a Type A quantity of 
radioactive material. Exception: this paragraph does not apply to a 
package--transported within the United States--containing greater than 
an A1 quantity (special form) of plutonium-244;
    (2) The package--transported within the United States--contains no 
more than 0.74 TBq (20 Ci) of special form plutonium-244; or
    (3) The package contains only LSA or SCO radioactive material, 
provided--
    (i) That the LSA or SCO material has an external radiation dose of 
less than or equal to 10 mSv/h (1 rem/h), at a distance of 3 m from the 
unshielded material; or
    (ii) That the package is classified as LSA-I or SCO-I.
    (c) A licensee is exempt from all the requirements of this part, 
other than Secs. 71.5 and 71.88, with respect to shipment or carriage 
of low-specific-activity (LSA) material in group LSA-I, or surface 
contaminated objects (SCOs) in group SCO-I.


Sec. 71.15  Exemption from classification as fissile material.

    Fissile materials meeting the requirements of at least one of the 
paragraphs (a) through (e) of this section are exempt from 
classification as fissile material and from the fissile material 
package standards of Secs. 71.55 and 71.59, but are subject to all 
other requirements of this part, except as noted.
    (a) The mass ratio of iron to fissile material is greater than 
200:1 and the package contents contain less than 15 g of fissile 
material. The fissile material may be contained in individual or bulk 
packaging.
    (b) The mass ratio of noncombustible, insoluble-in-water, material 
(including both the contents and packaging) to fissile material is 
greater than 2000:1 and the package contents contain less than 350 g of 
fissile material. Lead, beryllium, graphite, and hydrogenous material 
enriched in deuterium may be present in the package, but must not be 
included in determining the mass ratio for the package. The fissile 
material may be contained in individual or bulk packaging.
    (c) Uranium enriched in uranium-235 to a maximum of 1 percent by 
weight, and with total plutonium and uranium-233 content of up to 1 
percent of the mass of uranium-235, provided that the mass of any 
beryllium, graphite, and hydrogenous material enriched in deuterium 
present in the package is less than 0.1 percent of the fissile mass.
    (d) Liquid solutions of uranyl nitrate enriched in uranium-235 to a 
maximum of 2 percent by weight, provided that:
    (1) The total plutonium and uranium-233 content does not exceed 0.1 
percent of the mass of uranium-235;
    (2) The nitrogen to uranium atomic ratio (N/U) is greater than or 
equal to 2.0; and
    (3) The material must be contained in at least a DOT Type A 
package.
    (e) Plutonium with a total mass of less than 1000 grams, provided 
that: plutonium-239, plutonium-241, or any combination of these 
radionuclides, constitutes less than 20 percent by mass of the total 
quantity of plutonium in the package.


Sec. 71.16  [Reserved]

Subpart C--General Licenses


Sec. 71.17  General license: NRC-approved package.

    (a) A general license is hereby issued to any licensee of the 
Commission to transport, or to deliver to a carrier for transport, 
licensed material in a package (other than a Type B(DP) package) for 
which a license, certificate of compliance, or other approval has been 
issued by the NRC.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the certificate of compliance, or other approval 
of the package, and has the drawings and other documents referenced in 
the approval relating to the use and maintenance of the packaging and 
to the actions to be taken before shipment;
    (2) Complies with the terms and conditions of the license, 
certificate, or other approval, as applicable, and the applicable 
requirements of subparts A, G, and H of this part; and
    (3) Submits in writing to the NRC, before the licensee's first use 
of the package, the licensee's name and license number and the package 
identification number specified in the package approval. A licensee 
shall submit this information in accordance with Sec. 71.1.
    (d) This general license applies only when the package approval 
authorizes use of the package under this general license.
    (e) For a Type B or fissile material package, the design of which 
was approved by NRC before April 1, 1996, the general license is 
subject to the additional restrictions of Sec. 71.19.


Sec. 71.18  General license: NRC-approved Type B(DP) package.

    (a) A general license is hereby issued to any licensee of the 
Commission to

[[Page 21449]]

transport, or to deliver to a carrier for transport, licensed material 
in a Type B(DP) package for which a license, certificate of compliance 
(CoC), or other approval has been issued by the NRC.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the CoC, or other approval, of the Type B(DP) 
package, a copy of the updated final safety analysis report for the 
package, and the drawings and other documents referenced in the CoC, or 
other approval, relating to the use and maintenance of the packaging 
and to the actions to be taken before shipment;
    (2) Complies with the terms and conditions of the license, CoC, or 
other approval, as applicable, and the applicable requirements of 
subparts A, G, and H of this part; and
    (3) Submits in writing to the NRC, before the licensee's first use 
of the package, the licensee's name and license number and the package 
identification number specified in the package approval. A licensee 
shall submit this information in accordance with Sec. 71.1.
    (d) This general license applies only when the package approval 
authorizes use of the Type B(DP) package under this general license.
    (e) This general license does not authorize a Type B(DP) package to 
be transported by air.


Sec. 71.19  Previously approved package.

    (a) A Type B package previously approved by NRC, but not designated 
as B(U), B(M), B(U)F, B(M)F, in the identification number of the NRC 
Certificate of Compliance, or Type AF packages approved by the NRC 
prior to September 6, 1983, may be used under the general license of 
Sec. 71.17 until [date 3 years after the effective date of the final 
rule] with the following additional conditions:
    (1) Fabrication of the packaging was satisfactorily completed by 
August 31, 1986, as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A serial number that uniquely identifies each packaging which 
conforms to the approved design is assigned to, and legibly and durably 
marked on, the outside of each packaging; and
    (3) Sec. 71.19(a) will expire [date 3 years after the effective 
date of the final rule].
    (b) A Type B(U) package, a Type B(M) package, or a fissile material 
package, previously approved by the NRC but without the designation ``-
85'' in the identification number of the NRC Certificate of Compliance, 
may be used under the general license of Sec. 71.17 with the following 
additional conditions:
    (1) Fabrication of the package is satisfactorily completed by April 
1, 1999, as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A package used for a shipment to a location outside the United 
States is subject to multilateral approval as defined in DOT 
regulations at 49 CFR 173.403; and
    (3) A serial number which uniquely identifies each packaging which 
conforms to the approved design is assigned to and legibly and durably 
marked on the outside of each packaging.
    (c) A Type B(U) package, a Type B(M) package, or a fissile material 
package previously approved by the NRC, but without the designation ``-
85'' in the identification number of the NRC Certificate of Compliance, 
may be used under the general license of Sec. 71.17 with the following 
additional conditions:
    (1) Fabrication of the package must be satisfactorily completed by 
December 31, 2006, as demonstrated by application of its model number 
in accordance with Sec. 71.85(c); and
    (2) After December 31, 2003, a package used for a shipment to a 
location outside the United States is subject to multilateral approval 
as defined in DOT regulations at 49 CFR 173.403.
    (d) NRC will approve modifications to the design and authorized 
contents of a Type B package, or a fissile material package, previously 
approved by NRC, provided--
    (1) The modifications of a Type B package are not significant with 
respect to the design, operating characteristics, or safe performance 
of the containment system, when the package is subjected to the tests 
specified in Secs. 71.71 and 71.73;
    (2) The modifications of a fissile material package are not 
significant, with respect to the prevention of criticality, when the 
package is subjected to the tests specified in Secs. 71.71 and 71.73; 
and
    (3) The modifications to the package satisfy the requirements of 
this part.
    (e) NRC will revise the package identification number to designate 
previously approved package designs as B, BF, AF, B(U), B(M), B(U)F, 
B(M)F, B(U)-85, B(U)F-85, B(M)-85, B(M)F-85, or AF-85 as appropriate, 
and with the identification number suffix ``-96'' after receipt of an 
application demonstrating that the design meets the requirements of 
this part.


Sec. 71.20  General license: DOT specification container.

    (a) A general license is issued to any licensee of the Commission 
to transport, or to deliver to a carrier for transport, licensed 
material in a specification container for fissile material or for a 
Type B quantity of radioactive material as specified in DOT regulations 
at 49 CFR parts 173 and 178.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the specification; and
    (2) Complies with the terms and conditions of the specification and 
the applicable requirements of subparts A, G, and H of this part.
    (d) This general license is subject to the limitation that the 
specification container may not be used for a shipment to a location 
outside the United States, except by multilateral approval, as defined 
in DOT regulations at 49 CFR 173.403.


Sec. 71.21  General license: Use of foreign approved package.

    (a) A general license is issued to any licensee of the Commission 
to transport, or to deliver to a carrier for transport, licensed 
material in a package the design of which has been approved in a 
foreign national competent authority certificate that has been 
revalidated by DOT as meeting the applicable requirements of 49 CFR 
171.12.
    (b) Except as otherwise provided in this section, the general 
license applies only to a licensee who has a quality assurance program 
approved by the Commission as satisfying the applicable provisions of 
subpart H of this part.
    (c) This general license applies only to shipments made to or from 
locations outside the United States.
    (d) This general license applies only to a licensee who--
    (1) Has a copy of the applicable certificate, the revalidation, and 
the drawings and other documents referenced in the certificate, 
relating to the use and maintenance of the packaging and to the actions 
to be taken before shipment; and
    (2) Complies with the terms and conditions of the certificate and 
revalidation, and with the applicable requirements of subparts A, G, 
and H of this part. With respect to the quality assurance provisions of 
subpart H of

[[Page 21450]]

this part, the licensee is exempt from design, construction, and 
fabrication considerations.


Sec. 71.22  General license: Fissile material.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, if the material is shipped in accordance with 
this section. The fissile material need not be contained in a package 
which meets the standards of subparts E and F of this part; however, 
the material must be contained in a Type A package. The Type A package 
must also meet the DOT requirements of 49 CFR 173.417(a).
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) The general license applies only when a package's contents:
    (1) Contain less than a Type A quantity of fissile material; and
    (2) Contain less than 500 total grams of beryllium, graphite, or 
hydrogenous material enriched in deuterium.
    (d) The general license applies only to packages containing fissile 
material that are labeled with a CSI which:
    (1) Has been determined in accordance with paragraph (e) of this 
section;
    (2) Has a value less than or equal to 10.0; and
    (3) For a shipment of multiple packages containing fissile 
material, the sum of the CSIs must be less than or equal to 50.0 (for 
shipment on a nonexclusive use conveyance or storage incident to 
transport) and less than or equal to 100.0 (for shipment on an 
exclusive use conveyance).
    (e)(1) The value for the CSI must be greater than or equal to the 
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TP30AP02.034

    (2) The calculated CSI must be rounded up to the first decimal 
place;
    (3) The values of X, Y, and Z used in the CSI equation must be 
taken from Tables 71-1 or 71-2, as appropriate;
    (4) If Table 71-2 is used to obtain the value of X, then the values 
for the terms in the equation for uranium-233 and plutonium must be 
assumed to be zero; and
    (5) Table 71-1 values for X, Y, and Z must be used to determine the 
CSI if:
    (i) Uranium-233 is present in the package;
    (ii) The mass of plutonium exceeds 1 percent of the mass of 
uranium-235;
    (iii) The uranium-235 is of unknown enrichment; or
    (iv) Substances having a moderating effectiveness (i.e., an average 
hydrogen density greater than H2O) [e.g., certain 
hydrocarbon oils or plastics] are present in any form, except as 
polyethylene used for packing or wrapping.

Table 71-1.--Mass Limits for General License Packages Containing Mixed Quantities of Fissile Material or Uranium-
                                  235 of Unknown Enrichment per Sec.  71.22(e)
----------------------------------------------------------------------------------------------------------------
                                                                 Fissile material mass    Fissile material mass
                                                                 mixed with moderating    mixed with moderating
                                                                  substances having an     substances having an
                       Fissile material                             average hydrogen         average hydrogen
                                                                  density less than or     density greater than
                                                                 equal to H2O. (grams)        H2O a.(grams)
----------------------------------------------------------------------------------------------------------------
235U (X)......................................................                       60                       38
235U (Y)......................................................                       43                       27
239Pu or 241Pu (Z)............................................                       37                      24
----------------------------------------------------------------------------------------------------------------
aWhen mixtures of moderating substances are present, the lower mass limits shall be used if more than 15 percent
  of the moderating substance has an average hydrogen density greater than H2O.


Table 71-2.--Mass Limits for General License Packages Containing Uranium-
               235 of Known Enrichment per Sec.  71.22(e)
------------------------------------------------------------------------
                                                               Fissile
                                                               material
 Uranium enrichment in weight percent of 235U not exceeding    mass of
                                                              235U (X).
                                                               (grams)
------------------------------------------------------------------------
24.........................................................           60
20.........................................................           63
15.........................................................           67
11.........................................................           72
10.........................................................           76
9.5........................................................           78
9..........................................................           81
8.5........................................................           82
8..........................................................           85
7.5........................................................           88
7..........................................................           90
6.5........................................................           93
6..........................................................           97
5.5........................................................          102
5..........................................................          108
4.5........................................................          114
4..........................................................          120
3.5........................................................          132
3..........................................................          150
2.5........................................................          180
2..........................................................          246
1.5........................................................          408
1.35.......................................................          480
1..........................................................        1,020
0.92.......................................................        1,800
------------------------------------------------------------------------

Sec. 71.23  General license: Plutonium-beryllium special form material.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material in the form of plutonium-beryllium (Pu-
Be) special form sealed sources, or to deliver Pu-Be sealed sources to 
a carrier for transport, if the material is shipped in accordance

[[Page 21451]]

with this section. This material need not be contained in a package 
which meets the standards of subparts E and F of this part; however, 
the material must be contained in a Type A package. The Type A package 
must also meet the DOT requirements of 49 CFR 173.417(a).
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) The general license applies only when a package's contents:
    (1) Contain less than a Type A quantity of material; and
    (2) Contain less than 1000 g of plutonium, provided that: 
plutonium-239, plutonium-241, or any combination of these 
radionuclides, constitutes less than 240 g of the total quantity of 
plutonium in the package.
    (d) The general license applies only to packages labeled with a CSI 
which:
    (1) Has been determined in accordance with paragraph (e) of this 
section;
    (2) Has a value less than or equal to 100.0; and
    (3) For a shipment of multiple packages containing Pu-Be sealed 
sources, the sum of the CSIs must be less than or equal to 50.0 (for 
shipment on a nonexclusive use conveyance or storage incident to 
transport) and to less than or equal to 100.0 (for shipment on an 
exclusive use conveyance).
    (e)(1) The value for the CSI must be greater than or equal to the 
number calculated by the following equation:
[GRAPHIC] [TIFF OMITTED] TP30AP02.009

    (2) The calculated CSI must be rounded up to the first decimal 
place.


Sec. 71.24  [Reserved]


Sec. 71.25  [Reserved]

    3. In Sec. 71.41, paragraph (a) is revised and a new paragraph (d) 
is added to read as follows:


Sec. 71.41  Demonstration of compliance.

    (a) The effects on a package of the tests specified in Sec. 71.71 
(``Normal conditions of transport''), and the tests specified in 
Sec. 71.73 (``Hypothetical accident conditions''), and Sec. 71.61 
(``Special requirements for Type B packages containing more than 
105 A2''), must be evaluated by subjecting a 
specimen or scale model to a specific test, or by another method of 
demonstration acceptable to the Commission, as appropriate for the 
particular feature being considered.
* * * * *
    (d) Packages for which compliance with the other provisions of the 
regulations in this part is impracticable shall not be transported 
except under special package authorization. Provided the applicant 
demonstrates that compliance with the other provisions of the 
regulations is impracticable and that the requisite standards of safety 
established by these regulations have been demonstrated through means 
alternative to the other provisions, a special package authorization 
may be approved for one-time shipments. The applicant shall demonstrate 
that the overall level of safety in transport for these shipments is at 
least equivalent to that which would be provided if all the applicable 
requirements had been met.
    4. In Sec. 71.51, the introductory text of paragraph (a) is 
revised, and a new paragraph (d) is added to read as follows:


Sec. 71.51  Additional requirements for Type B packages.

    (a) A Type B package, in addition to satisfying the requirements of 
Secs. 71.41 through 71.47, must be designed, constructed, and prepared 
for shipment so that under the tests specified in:
* * * * *
    (d) For packages which contain radioactive contents with activity 
greater than 105 A2, the requirements of 
Sec. 71.61 must be met. This requirement does not apply to Type B(DP) 
packages.


Sec. 71.53  [Reserved]

    5. Section 71.53 is removed and reserved.
    6. In Sec. 71.55, paragraph (b) introductory text is revised, and 
new paragraphs (f) and (g) are added to read as follows:


Sec. 71.55  General requirements for fissile material packages.

* * * * *
    (b) Except as provided in paragraph (c) or (g) of this section, a 
package used for the shipment of fissile material must be so designed 
and constructed and its contents so limited that it would be 
subcritical if water were to leak into the containment system, or 
liquid contents were to leak out of the containment system so that, 
under the following conditions, maximum reactivity of the fissile 
material would be attained:
* * * * *
    (f) For fissile material package designs to be transported by air:
    (1) The package must be designed and constructed, and its contents 
limited so that it would be subcritical, assuming reflection by 20 cm 
(7.9 in) of water but no water inleakage, when subjected to sequential 
application of:
    (i) The free drop test in Sec. 71.73(c)(1);
    (ii) The crush test in Sec. 71.73(c)(2);
    (iii) A puncture test, for packages of 250 kg or more, consisting 
of a free drop of the specimen through a distance of 3 m (120 in) in a 
position for which maximum damage is expected at the conclusion of the 
test sequence, onto the upper end of a solid, vertical, cylindrical, 
mild steel probe mounted on an essentially unyielding, horizontal 
surface. The probe must be 20 cm (7.9 in) in diameter, with the 
striking end forming the frustum of a right circular cone with the 
dimensions of 30 cm height, 2.5 cm top diameter, and a top edge rounded 
to a radius of not more than 6 mm (0.25 in). For packages less than 250 
kg, the puncture test must be the same, except that a 250 kg probe must 
be dropped onto the specimen which must be placed on the surface; and
    (iv) The thermal test in Sec. 71.73(c)(4), except that the duration 
of the test must be 60 minutes.
    (2) The package must be designed and constructed, and its contents 
limited so that it would be subcritical, assuming reflection by 20 cm 
(7.9 in) of water but no water inleakage, when subjected to an impact 
on an unyielding surface at a velocity of 90 m/s normal to the surface, 
at such orientation so as to result in maximum damage. A separate, 
undamaged specimen can be used for this evaluation.
    (3) Allowance may not be made for the special design features in 
paragraph (c) of this section, unless water leakage into or out of void 
spaces is prevented following application of the tests in paragraphs 
(f)(1) and (f)(2) of this section, and subsequent application of the 
immersion test in Sec. 71.73(c)(5).
    (g) Packages containing uranium hexafluoride only are excepted from 
the requirements of paragraph (b) of this section provided that:
    (1) Following the tests specified in Sec. 71.73 (``Hypothetical 
accident conditions''), there is no physical

[[Page 21452]]

contact between the valve body and any other component of the 
packaging, other than at its original point of attachment, and the 
valve remains leak tight;
    (2) There is an adequate quality control in the manufacture, 
maintenance, and repair of packagings;
    (3) Each package is tested to demonstrate closure before each 
shipment; and
    (4) The uranium is enriched to not more than 5 weight percent 
uranium-235.
    7. In Sec. 71.59, paragraphs (b) and (c) are revised to read as 
follows:


Sec. 71.59  Standards for arrays of fissile material packages.

* * * * *
    (b) The CSI must be determined by dividing the number 50 by the 
value of ``N'' derived using the procedures specified in paragraph (a) 
of this section. The value of the CSI may be zero provided that an 
unlimited number of packages are subcritical, such that the value of 
``N'' is effectively equal to infinity under the procedures specified 
in paragraph (a) of this section. Any CSI greater than zero must be 
rounded up to the first decimal place.
    (c) For a fissile material package which is assigned a CSI value--
    (1) Less than or equal to 50.0, that package may be shipped by a 
carrier in a nonexclusive use conveyance, or stored incident to 
transport, provided the sum of the CSIs is limited to less than or 
equal to 50.0.
    (2) Less than or equal to 50.0, that package may be shipped by a 
carrier in an exclusive use conveyance, provided the sum of the CSIs is 
limited to less than or equal to 100.0.
    (3) Greater than 50.0, that package must be shipped by a carrier in 
an exclusive use conveyance, provided the sum of the CSIs is limited to 
less than or equal to 100.0.
    8. Section 71.61 is revised to read as follows:


Sec. 71.61  Special requirements for Type B packages containing more 
than 10\5\A2.

    A Type B package containing more than 10\5\ A2 must be 
designed so that its undamaged containment system can withstand an 
external water pressure of 2 MPa (290 psi) for a period of not less 
than 1 hour without collapse, buckling, or inleakage of water.
    9. Section 71.63 is revised to read as follows:


Sec. 71.63  Special requirement for plutonium shipments.

    Shipments containing plutonium must be made with the contents in 
solid form, if the contents contain greater than 0.74 TBq (20 Ci) of 
plutonium.
    10. In Sec. 71.73, paragraph (c)(2) is revised to read as follows:


Sec. 71.73  Hypothetical accident conditions.

* * * * *
    (c)* * *
    (2) Crush. Subjection of the specimen to a dynamic crush test by 
positioning the specimen on a flat, essentially unyielding horizontal 
surface so as to suffer maximum damage by the drop of a 500-kg (1100-
lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a 
solid mild steel plate 1 m (40 in) by 1 m and must fall in a horizontal 
attitude. The crush test is required only when the specimen has a mass 
not greater than 500 kg (1100 lbs), an overall density not greater than 
1000 kg/m\3\ (62.4 lbs/ft\3\) based on external dimension, and 
radioactive contents greater than 1000 A\2\ not as special form 
radioactive material. For packages containing fissile material, the 
radioactive contents greater than 1000 A\2\ criterion does not apply.
* * * * *
    11. In Sec. 71.88, paragraph (a)(2) is revised to read as follows:


Sec. 71.88  Air transport of plutonium.

    (a) * * *
    (2) The plutonium is contained in a material in which the specific 
activity is less than or equal to the activity concentration values for 
plutonium specified in Appendix A, Table A-2 of this part, and in which 
the radioactivity is essentially uniformly distributed; or
* * * * *
    12. In Sec. 71.91, paragraphs (b) and (c) are revised, and a new 
paragraph (d) is added to read as follows:


Sec. 71.91  Records.

* * * * *
    (b) Each certificate holder shall maintain, for a period of 3 years 
after the life of the packaging to which they apply, records 
identifying the packaging by model number, serial number, and date of 
manufacture.
    (c) The licensee, certificate holder, and an applicant for a CoC, 
shall make available to the Commission for inspection, upon reasonable 
notice, all records required by this part. Records are only valid if 
stamped, initialed, or signed and dated by authorized personnel or 
otherwise authenticated.
    (d) The licensee, certificate holder, and an applicant for a CoC 
shall maintain sufficient written records to furnish evidence of the 
quality of packaging. The records to be maintained include results of 
the determinations required by Sec. 71.85; design, fabrication, and 
assembly records, results of reviews, inspections, tests, and audits; 
results of monitoring work performance and materials analyses; and 
results of maintenance, modification, and repair activities. 
Inspection, test, and audit records must identify the inspector or data 
recorder, the type of observation, the results, the acceptability, and 
the action taken in connection with any deficiencies noted. These 
records must be retained for 3 years after the life of the packaging to 
which they apply.
    13. Section 71.93 is revised to read as follows:


Sec. 71.93  Inspection and tests.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
permit the Commission, at all reasonable times, to inspect the licensed 
material, packaging, premises, and facilities in which the licensed 
material or packaging is used, provided, constructed, fabricated, 
tested, stored, or shipped.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
perform, and permit the Commission to perform, any tests the Commission 
deems necessary or appropriate for the administration of the 
regulations in this chapter.
    (c) The certificate holder and applicant for a CoC shall notify the 
NRC, in accordance with Sec. 71.1, 45 days in advance of starting 
fabrication of the first packaging under a CoC. This paragraph applies 
to any packaging used for the shipment of licensed material which has 
either--
    (1) A decay heat load in excess of 5 kW; or
    (2) A maximum normal operating pressure in excess of 103 kPa (15 
lbf/in \2\) gauge.
    14. Section 71.95 is revised to read as follows:


Sec. 71.95  Reports.

    (a) The licensee, after requesting the certificate holder's input, 
shall submit a written report to the Commission of'
    (1) Instances in which there is a significant reduction in the 
effectiveness of any NRC-approved Type B or Type A(F) packaging during 
use; or
    (2) Details of any defects with safety significance in any NRC-
approved Type B or fissile material packaging, after first use.
    (b) The licensee shall submit a written report to the Commission of 
instances in which the conditions in the certificate of compliance were 
not followed during a shipment.
    (c) Written report. Each licensee shall submit, in accordance with 
Sec. 71.1, a written report required by paragraph (a) or (b) of this 
section within 60 days of the event or discovery of the event. The 
licensee shall also provide a copy of

[[Page 21453]]

each report submitted to the NRC to the applicable certificate holder. 
Written reports prepared pursuant to other regulations may be submitted 
to fulfill this requirement if the reports contain all the necessary 
information, and the appropriate distribution is made. These written 
reports must include the following:
    (1) A brief abstract describing the major occurrences during the 
event, including all component or system failures that contributed to 
the event and significant corrective action taken or planned to prevent 
recurrence.
    (2) A clear, specific, narrative description of the event that 
occurred so that knowledgeable readers conversant with the requirements 
of Part 71, but not familiar with the design of the packaging, can 
understand the complete event. The narrative description must include 
the following specific information as appropriate for the particular 
event.
    (i) Status of components or systems that were inoperable at the 
start of the event and that contributed to the event;
    (ii) Dates and approximate times of occurrences;
    (iii) The cause of each component or system failure or personnel 
error, if known;
    (iv) The failure mode, mechanism, and effect of each failed 
component, if known;
    (v) A list of systems or secondary functions that were also 
affected for failures of components with multiple functions;
    (vi) The method of discovery of each component or system failure or 
procedural error;
    (vii) For each human performance-related root cause, a discussion 
of the cause(s) and circumstances; (viii) The manufacturer and model 
number (or other identification) of each component that failed during 
the event; and
    (ix) For events occurring during use of a packaging, the quantities 
and chemical and physical form(s) of the package contents.
    (3) An assessment of the safety consequences and implications of 
the event. This assessment must include the availability of other 
systems or components that could have performed the same function as 
the components and systems that failed during the event.
    (4) A description of any corrective actions planned as a result of 
the event, including the means employed to repair any defects, and 
actions taken to reduce the probability of similar events occurring in 
the future.
    (5) Reference to any previous similar events involving the same 
packaging that are known to the licensee or certificate holder.
    (6) The name and telephone number of a person within the licensee's 
organization who is knowledgeable about the event and can provide 
additional information.
    (7) The extent of exposure of individuals to radiation or to 
radioactive materials without identification of individuals by name.
    (d) Report legibility. The reports submitted by licensees and/or 
certificate holders under this section must be of sufficient quality to 
permit reproduction and micrographic processing.
    15. In Sec. 71.100, paragraph (b) is revised to read as follows:


Sec. 71.100  Criminal penalties.

* * * * *
    (b) The regulations in part 71 that are not issued under sections 
161b, 161i, or 161o for the purposes of section 223 are as follows: 
Secs. 71.0, 71.2, 71.4, 71.6, 71.7, 71.10, 71.31, 71.33, 71.35, 71.37, 
71.38, 71.39, 71.40, 71.41, 71.43, 71.45, 71.47, 71.51, 71.55, 71.59, 
71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, 71.100, and 71.151 
through 71.169.
    16. Subpart H to Part 71 is revised to read as follows:
Subpart H--Quality Assurance
Sec.
71.101   Quality assurance requirements.
71.103   Quality assurance organization.
71.105   Quality assurance program.
71.107   Package design control.
71.109   Procurement document control.
71.111   Instructions, procedures, and drawings.
71.113   Document control.
71.115   Control of purchased material, equipment, and services.
71.117   Identification and control of materials, parts, and 
components.
71.119   Control of special processes.
71.121   Internal inspection.
71.123   Test control.
71.125   Control of measuring and test equipment.
71.127   Handling, storage, and shipping control.
71.129   Inspection, test, and operating status.
71.131   Nonconforming materials, parts, or components.
71.133   Corrective action.
71.135   Quality assurance records.
71.137   Audits.

Subpart H--Quality Assurance


Sec. 71.101  Quality assurance requirements.

    (a) Purpose. This subpart describes quality assurance requirements 
applying to design, purchase, fabrication, handling, shipping, storing, 
cleaning, assembly, inspection, testing, operation, maintenance, 
repair, and modification of components of packaging that are important 
to safety. As used in this subpart, ``quality assurance'' comprises all 
those planned and systematic actions necessary to provide adequate 
confidence that a system or component will perform satisfactorily in 
service. Quality assurance includes quality control, which comprises 
those quality assurance actions related to control of the physical 
characteristics and quality of the material or component to 
predetermined requirements. The licensee, certificate holder, and 
applicant for a CoC are responsible for the quality assurance 
requirements as they apply to design, fabrication, testing, and 
modification of packaging. Each licensee is responsible for the quality 
assurance provision which applies to its use of a packaging for the 
shipment of licensed material subject to this subpart.
    (b) Establishment of program. Each licensee, certificate holder, 
and applicant for a CoC shall establish, maintain, and execute a 
quality assurance program satisfying each of the applicable criteria of 
Secs. 71.101 through 71.137 and satisfying any specific provisions that 
are applicable to the licensee's activities including procurement of 
packaging. The licensee, certificate holder, and applicant for a CoC 
shall execute the applicable criteria in a graded approach to an extent 
that is commensurate with the quality assurance requirement's 
importance to safety.
    (c) Approval of program. (1) Before the use of any package for the 
shipment of licensed material subject to this subpart, each licensee 
shall obtain Commission approval of its quality assurance program. Each 
licensee shall, in accordance with Sec. 71.1, file a description of its 
quality assurance program, including a discussion of which requirements 
of this subpart are applicable and how they will be satisfied.
    (2) Before the fabrication, testing, or modification of any package 
for the shipment of licensed material subject to this subpart, each 
licensee, certificate holder, or applicant for a CoC shall obtain 
Commission approval of its quality assurance program. Each certificate 
holder or applicant for a CoC shall, in accordance with Sec. 71.1, file 
a description of its quality assurance program, including a discussion 
of which requirements of this subpart are applicable and how they will 
be satisfied.
    (d) Existing package designs. The provisions of this paragraph deal 
with packages that have been approved for use in accordance with this 
part before January 1, 1979, and which have been

[[Page 21454]]

designed in accordance with the provisions of this part in effect at 
the time of application for package approval. Those packages will be 
accepted as having been designed in accordance with a quality assurance 
program that satisfies the provisions of paragraph (b) of this section.
    (e) Existing packages. The provisions of this paragraph deal with 
packages that have been approved for use in accordance with this part 
before January 1, 1979, have been at least partially fabricated before 
that date, and for which the fabrication is in accordance with the 
provisions of this part in effect at the time of application for 
approval of package design. These packages will be accepted as having 
been fabricated and assembled in accordance with a quality assurance 
program that satisfies the provisions of paragraph (b) of this section.
    (f) Previously approved programs. A Commission-approved quality 
assurance program that satisfies the applicable criteria of subpart H 
of this part, Appendix B of part 50 of this chapter, or subpart G of 
part 72 of this chapter, and that is established, maintained, and 
executed regarding transport packages, will be accepted as satisfying 
the requirements of paragraph (b) of this section. Before first use, 
the licensee, certificate holder, and applicant for a CoC shall notify 
the NRC, in accordance with Sec. 71.1, of its intent to apply its 
previously approved subpart H, Appendix B, or subpart G quality 
assurance program to transportation activities. The licensee, 
certificate holder, and applicant for a CoC shall identify the program 
by date of submittal to the Commission, Docket Number, and date of 
Commission approval.
    (g) Radiography containers. A program for transport container 
inspection and maintenance limited to radiographic exposure devices, 
source changers, or packages transporting these devices and meeting the 
requirements of Sec. 34.31(b) of this chapter or equivalent Agreement 
State requirement, is deemed to satisfy the requirements of 
Secs. 71.17(b) and 71.101(b).


Sec. 71.103  Quality assurance organization.

    (a) The licensee, \2\ certificate holder, and applicant for a CoC 
shall be responsible for the establishment and execution of the quality 
assurance program. The licensee, certificate holder, and applicant for 
a CoC may delegate to others, such as contractors, agents, or 
consultants, the work of establishing and executing the quality 
assurance program, or any part of the quality assurance program, but 
shall retain responsibility for the program. The licensee, certificate 
holder, and applicant for a CoC shall clearly establish and delineate, 
in writing, the authority and duties of persons and organizations 
performing activities affecting the functions of structures, systems, 
and components that are important to safety. These activities include 
performing the functions associated with attaining quality objectives 
and the quality assurance functions.
---------------------------------------------------------------------------

    \2\ While the term ``licensee'' is used in these criteria, the 
requirements are applicable to whatever design, fabrication, 
assembly, and testing of the package is accomplished with respect to 
a package before the time a package approval is issued.
---------------------------------------------------------------------------

    (b) The quality assurance functions are--
    (1) Assuring that an appropriate quality assurance program is 
established and effectively executed; and
    (2) Verifying, by procedures such as checking, auditing, and 
inspection, that activities affecting the functions that are important 
to safety have been correctly performed.
    (c) The persons and organizations performing quality assurance 
functions must have sufficient authority and organizational freedom 
to--
    (1) Identify quality problems;
    (2) Initiate, recommend, or provide solutions; and
    (3) Verify implementation of solutions.
    (d) The persons and organizations performing quality assurance 
functions shall report to a management level that assures that the 
required authority and organizational freedom, including sufficient 
independence from cost and schedule, when opposed to safety 
considerations, are provided.
    (e) Because of the many variables involved, such as the number of 
personnel, the type of activity being performed, and the location or 
locations where activities are performed, the organizational structure 
for executing the quality assurance program may take various forms, 
provided that the persons and organizations assigned the quality 
assurance functions have the required authority and organizational 
freedom.
    (f) Irrespective of the organizational structure, the individual(s) 
assigned the responsibility for assuring effective execution of any 
portion of the quality assurance program, at any location where 
activities subject to this section are being performed, must have 
direct access to the levels of management necessary to perform this 
function.


Sec. 71.105  Quality assurance program.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish, at the earliest practicable time consistent with the 
schedule for accomplishing the activities, a quality assurance program 
that complies with the requirements of Secs. 71.101 through 71.137. The 
licensee, certificate holder, and applicant for a CoC shall document 
the quality assurance program by written procedures or instructions and 
shall carry out the program in accordance with those procedures 
throughout the period during which the packaging is used. The licensee, 
certificate holder, and applicant for a CoC shall identify the material 
and components to be covered by the quality assurance program, the 
major organizations participating in the program, and the designated 
functions of these organizations.
    (b) The licensee, certificate holder, and applicant for a CoC, 
through its quality assurance program, shall provide control over 
activities affecting the quality of the identified materials and 
components to an extent consistent with their importance to safety, and 
as necessary to assure conformance to the approved design of each 
individual package used for the shipment of radioactive material. The 
licensee, certificate holder, and applicant for a CoC shall assure that 
activities affecting quality are accomplished under suitably controlled 
conditions. Controlled conditions include the use of appropriate 
equipment; suitable environmental conditions for accomplishing the 
activity, such as adequate cleanliness; and assurance that all 
prerequisites for the given activity have been satisfied. The licensee, 
certificate holder, and applicant for a CoC shall take into account the 
need for special controls, processes, test equipment, tools, and skills 
to attain the required quality, and the need for verification of 
quality by inspection and test.
    (c) The licensee, certificate holder, and applicant for a CoC shall 
base the requirements and procedures of its quality assurance program 
on the following considerations concerning the complexity and proposed 
use of the package and its components:
    (1) The impact of malfunction or failure of the item to safety;
    (2) The design and fabrication complexity or uniqueness of the 
item;
    (3) The need for special controls and surveillance over processes 
and equipment;
    (4) The degree to which functional compliance can be demonstrated 
by inspection or test; and
    (5) The quality history and degree of standardization of the item.

[[Page 21455]]

    (d) The licensee, certificate holder, and applicant for a CoC shall 
provide for indoctrination and training of personnel performing 
activities affecting quality, as necessary to assure that suitable 
proficiency is achieved and maintained. The licensee, certificate 
holder, and applicant for a CoC shall review the status and adequacy of 
the quality assurance program at established intervals. Management of 
other organizations participating in the quality assurance program 
shall review regularly the status and adequacy of that part of the 
quality assurance program they are executing.


Sec. 71.107  Package design control.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that applicable regulatory requirements 
and the package design, as specified in the license or CoC for those 
materials and components to which this section applies, are correctly 
translated into specifications, drawings, procedures, and instructions. 
These measures must include provisions to assure that appropriate 
quality standards are specified and included in design documents and 
that deviations from standards are controlled. Measures must be 
established for the selection and review for suitability of application 
of materials, parts, equipment, and processes that are essential to the 
functions of the materials, parts, and components of the packaging that 
are important to safety.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
establish measures for the identification and control of design 
interfaces and for coordination among participating design 
organizations. These measures must include the establishment of written 
procedures, among participating design organizations, for the review, 
approval, release, distribution, and revision of documents involving 
design interfaces. The design control measures must provide for 
verifying or checking the adequacy of design, by methods such as design 
reviews, alternate or simplified calculational methods, or by a 
suitable testing program. For the verifying or checking process, the 
licensee shall designate individuals or groups other than those who 
were responsible for the original design, but who may be from the same 
organization. Where a test program is used to verify the adequacy of a 
specific design feature in lieu of other verifying or checking 
processes, the licensee, certificate holder, and applicant for a CoC 
shall include suitable qualification testing of a prototype or sample 
unit under the most adverse design conditions. The licensee, 
certificate holder, and applicant for a CoC shall apply design control 
measures to the following:

    (1) Criticality physics, radiation shielding, stress, thermal, 
hydraulic, and accident analyses;
    (2) Compatibility of materials;
    (3) Accessibility for inservice inspection, maintenance, and 
repair;
    (4) Features to facilitate decontamination; and
    (5) Delineation of acceptance criteria for inspections and 
tests.

    (c) The licensee, certificate holder, and applicant for a CoC shall 
subject design changes, including field changes, to design control 
measures commensurate with those applied to the original design. 
Changes in the conditions specified in the CoC require NRC prior 
approval.


Sec. 71.109  Procurement document control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that adequate quality is required in the 
documents for procurement of material, equipment, and services, whether 
purchased by the licensee, certificate holder, and applicant for a CoC 
or by its contractors or subcontractors. To the extent necessary, the 
licensee, certificate holder, and applicant for a CoC shall require 
contractors or subcontractors to provide a quality assurance program 
consistent with the applicable provisions of this part.


Sec. 71.111  Instructions, procedures, and drawings.

    The licensee, certificate holder, and applicant for a CoC shall 
prescribe activities affecting quality by documented instructions, 
procedures, or drawings of a type appropriate to the circumstances and 
shall require that these instructions, procedures, and drawings be 
followed. The instructions, procedures, and drawings must include 
appropriate quantitative or qualitative acceptance criteria for 
determining that important activities have been satisfactorily 
accomplished.


Sec. 71.113  Document control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control the issuance of documents such as 
instructions, procedures, and drawings, including changes, that 
prescribe all activities affecting quality. These measures must assure 
that documents, including changes, are reviewed for adequacy, approved 
for release by authorized personnel, and distributed and used at the 
location where the prescribed activity is performed.


Sec. 71.115  Control of purchased material, equipment, and services.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that purchased material, equipment, and 
services, whether purchased directly or through contractors and 
subcontractors, conform to the procurement documents. These measures 
must include provisions, as appropriate, for source evaluation and 
selection, objective evidence of quality furnished by the contractor or 
subcontractor, inspection at the contractor or subcontractor source, 
and examination of products on delivery.
    (b) The licensee, certificate holder, and applicant for a CoC shall 
have available documentary evidence that material and equipment conform 
to the procurement specifications before installation or use of the 
material and equipment. The licensee, certificate holder, and applicant 
for a CoC shall retain, or have available, this documentary evidence 
for the life of the package to which it applies. The licensee, 
certificate holder, and applicant for a CoC shall assure that the 
evidence is sufficient to identify the specific requirements met by the 
purchased material and equipment.
    (c) The licensee, certificate holder, and applicant for a CoC shall 
assess the effectiveness of the control of quality by contractors and 
subcontractors at intervals consistent with the importance, complexity, 
and quantity of the product or services.


Sec. 71.117  Identification and control of materials, parts, and 
components.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures for the identification and control of materials, 
parts, and components. These measures must assure that identification 
of the item is maintained by heat number, part number, or other 
appropriate means, either on the item or on records traceable to the 
item, as required throughout fabrication, installation, and use of the 
item. These identification and control measures must be designed to 
prevent the use of incorrect or defective materials, parts, and 
components.


Sec. 71.119  Control of special processes.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that special processes, including welding, 
heat treating, and nondestructive testing, are controlled and 
accomplished by qualified personnel using qualified procedures in 
accordance with applicable codes, standards,

[[Page 21456]]

specifications, criteria, and other special requirements.


Sec. 71.121  Internal inspection.

    The licensee, certificate holder, and applicant for a CoC shall 
establish and execute a program for inspection of activities affecting 
quality by or for the organization performing the activity, to verify 
conformance with the documented instructions, procedures, and drawings 
for accomplishing the activity. The inspection must be performed by 
individuals other than those who performed the activity being 
inspected. Examination, measurements, or tests of material or products 
processed must be performed for each work operation where necessary to 
assure quality. If direct inspection of processed material or products 
is not carried out, indirect control by monitoring processing methods, 
equipment, and personnel must be provided. Both inspection and process 
monitoring must be provided when quality control is inadequate without 
both. If mandatory inspection hold points, which require witnessing or 
inspecting by the licensee's designated representative and beyond which 
work should not proceed without the consent of its designated 
representative, are required, the specific hold points must be 
indicated in appropriate documents.


Sec. 71.123  Test control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish a test program to assure that all testing required to 
demonstrate that the packaging components will perform satisfactorily 
in service is identified and performed in accordance with written test 
procedures that incorporate the requirements of this part and the 
requirements and acceptance limits contained in the package approval. 
The test procedures must include provisions for assuring that all 
prerequisites for the given test are met, that adequate test 
instrumentation is available and used, and that the test is performed 
under suitable environmental conditions. The licensee, certificate 
holder, and applicant for a CoC shall document and evaluate the test 
results to assure that test requirements have been satisfied.


Sec. 71.125  Control of measuring and test equipment.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that tools, gauges, instruments, and other 
measuring and testing devices used in activities affecting quality are 
properly controlled, calibrated, and adjusted at specified times to 
maintain accuracy within necessary limits.


Sec. 71.127  Handling, storage, and shipping control.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control, in accordance with instructions, the 
handling, storage, shipping, cleaning, and preservation of materials 
and equipment to be used in packaging to prevent damage or 
deterioration. When necessary for particular products, special 
protective environments, such as inert gas atmosphere, and specific 
moisture content and temperature levels must be specified and provided.


Sec. 71.129  Inspection, test, and operating status.

    (a) The licensee, certificate holder, and applicant for a CoC shall 
establish measures to indicate, by the use of markings such as stamps, 
tags, labels, routing cards, or other suitable means, the status of 
inspections and tests performed upon individual items of the packaging. 
These measures must provide for the identification of items that have 
satisfactorily passed required inspections and tests, where necessary 
to preclude inadvertent bypassing of the inspections and tests.
    (b) The licensee shall establish measures to identify the operating 
status of components of the packaging, such as tagging valves and 
switches, to prevent inadvertent operation.


Sec. 71.131  Nonconforming materials, parts, or components.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to control materials, parts, or components that do 
not conform to the licensee's requirements to prevent their inadvertent 
use or installation. These measures must include, as appropriate, 
procedures for identification, documentation, segregation, disposition, 
and notification to affected organizations. Nonconforming items must be 
reviewed and accepted, rejected, repaired, or reworked in accordance 
with documented procedures.


Sec. 71.133  Corrective action.

    The licensee, certificate holder, and applicant for a CoC shall 
establish measures to assure that conditions adverse to quality, such 
as deficiencies, deviations, defective material and equipment, and 
nonconformances, are promptly identified and corrected. In the case of 
a significant condition adverse to quality, the measures must assure 
that the cause of the condition is determined and corrective action 
taken to preclude repetition. The identification of the significant 
condition adverse to quality, the cause of the condition, and the 
corrective action taken must be documented and reported to appropriate 
levels of management.


Sec. 71.135  Quality assurance records.

    The licensee, certificate holder, and applicant for a CoC shall 
maintain sufficient written records to describe the activities 
affecting quality. The records must include the instructions, 
procedures, and drawings required by Sec. 71.111 to prescribe quality 
assurance activities and must include closely related specifications 
such as required qualifications of personnel, procedures, and 
equipment. The records must include the instructions or procedures 
which establish a records retention program that is consistent with 
applicable regulations and designates factors such as duration, 
location, and assigned responsibility. The licensee, certificate 
holder, and applicant for a CoC shall retain these records for 3 years 
beyond the date when the licensee, certificate holder, and applicant 
for a CoC last engage in the activity for which the quality assurance 
program was developed. If any portion of the written procedures or 
instructions is superseded, the licensee, certificate holder, and 
applicant for a CoC shall retain the superseded material for 3 years 
after it is superseded.


Sec. 71.137  Audits.

    The licensee, certificate holder, and applicant for a CoC shall 
carry out a comprehensive system of planned and periodic audits, to 
verify compliance with all aspects of the quality assurance program, 
and to determine the effectiveness of the program. The audits must be 
performed in accordance with written procedures or checklists by 
appropriately trained personnel not having direct responsibilities in 
the areas being audited. Audited results must be documented and 
reviewed by management having responsibility in the area audited. 
Follow-up action, including reaudit of deficient areas, must be taken 
where indicated.
    17. A new subpart I is added to Part 71 to read as follows:
Subpart I--Type B(DP) Package Approval
Sec.
71.151   Procedures for applying for a Type B(DP) package approval.
71.153   Contents of application.
71.155   Package description.
71.157   Package evaluation.
71.159   Quality assurance.
71.161   Requirement for additional information.

[[Page 21457]]

71.163   Issuance of an NRC certificate of compliance.
71.165  Conditions for package reapproval.
71.167  Application to amend a certificate of compliance.
71.169  Issuance of an amendment to a certificate of compliance.
71.171  Inspections and tests.
71.173  Recordkeeping and reports.
71.175  Changes.
71.177  Safety analysis report updating.

Subpart I--Type B(DP) Package Approval


Sec. 71.151  Procedures for applying for a Type B(DP) package approval.

    (a) Spent fuel storage casks that have been issued a Certificate of 
Compliance (CoC) under subpart L of part 72 of this chapter may also be 
approved under this subpart as a Type B(DP) package for the 
transportation of spent fuel. A copy of the part 72 CoC issued for the 
cask, and any drawings and other documents referenced in the part 72 
CoC, must be included with the application.
    (b) An application for approval of a Type B(DP) package design must 
contain the information required by Sec. 71.153 and be submitted in 
accordance with Sec. 71.1.
    (c) Public inspection. An application for the approval of a Type 
B(DP) package, or amendment of a Type B(DP) package, may be made 
available for public inspection under Sec. 71.10.
    (d) Fees. Fees for reviews and evaluations related to issuance of a 
Type B(DP) CoC and inspections related to package fabrication are those 
shown in Sec. 170.31 of this chapter.


Sec. 71.153  Contents of application.

    (a) An application for an approval of a Type B(DP) package under 
this subpart must include, for each proposed Type B(DP) packaging 
design, the following information:
    (1) A package description as required by Sec. 71.155;
    (2) A package evaluation as required by Sec. 71.157; and
    (3) A quality assurance program description, as required by 
Sec. 71.159, or a reference to a previously approved quality assurance 
program.
    (b) A safety analysis report describing--
    (1) The proposed Type B(DP) package design;
    (2) How the package would be used to transport spent fuel safely;
    (3) An analysis of potential accidents, package response to these 
potential accidents, and any consequences to the public; and
    (4) How the package is suitable for the transportation of spent 
fuel for a period of at least 20 years.
    (c) Except as provided in Sec. 71.19, an application for 
modification of a Type B(DP) package design, whether for modification 
of the packaging or the authorized contents, must include sufficient 
information to demonstrate that the proposed design satisfies the Type 
B(DP) package standards in effect at the time the application is filed.
    (d) The applicant shall identify any established codes and 
standards proposed for use in package design, fabrication, assembly, 
testing, maintenance, and use. In the absence of any codes and 
standards, the applicant shall describe and justify the basis and 
rationale used to formulate the package quality assurance program.


Sec. 71.155  Package description.

    The application must include a description of the proposed Type 
B(DP) package in sufficient detail to identify the Type B(DP) package 
accurately and provide a sufficient basis for evaluation of the Type 
B(DP) package. The description must include--
    (a) With respect to the packaging--
    (1) Gross weight;
    (2) Model number;
    (3) Identification of the containment system;
    (4) Specific materials of construction, weights, dimensions, and 
fabrication methods of--
    (i) Receptacles;
    (ii) Materials specifically used as nonfissile neutron absorbers or 
moderators;
    (iii) Internal and external structures supporting or protecting 
receptacles;
    (iv) Valves, sampling ports, lifting devices, and tie-down devices; 
and
    (v) Structural and mechanical means for the transfer and 
dissipation of heat; and
    (5) Identification and volumes of any receptacles containing 
coolant.
    (b) With respect to the contents of the package--
    (1) Identification and maximum radioactivity of radioactive 
constituents;
    (2) Identification and maximum quantities of fissile constituents;
    (3) Chemical and physical form;
    (4) Extent of reflection, the amount and identity of nonfissile 
materials used as neutron absorbers or moderators, and the atomic ratio 
of moderator to fissile constituents;
    (5) Maximum normal operating pressure;
    (6) Maximum weight;
    (7) Maximum amount of decay heat; and
    (8) Identification and volumes of any coolants.


Sec. 71.157  Package evaluation.

    The application submitted under Sec. 71.151 must include the 
following:
    (a) A demonstration that the Type B(DP) package satisfies the 
standards specified in subparts E and F of this part. The application 
need not address the requirements of Secs. 71.61, 71.64, 71.74, 71.75, 
and 71.77;
    (b) The number ``N'' for the Type B(DP) package as determined in 
accordance with Sec. 71.59; and
    (c) Any proposed special controls and precautions for transport, 
loading, unloading, and handling, and any proposed special controls in 
case of an accident or delay.


Sec. 71.159  Quality assurance.

    (a) The applicant shall describe the quality assurance program (see 
subpart H of this part) for the design, fabrication, assembly, testing, 
maintenance, repair, modification, and use of the proposed Type B(DP) 
package.
    (b) The applicant shall identify any specific provisions of the 
quality assurance program that are applicable to the particular Type 
B(DP) package design under consideration, including a description of 
any leak testing.


Sec. 71.161  Requirement for additional information.

    The Commission may at any time require additional information to 
enable it to determine whether a license, CoC, or other approval should 
be granted, renewed, denied, modified, suspended, or revoked.


Sec. 71.163  Issuance of an NRC certificate of compliance.

    The NRC will issue a CoC for a Type B(DP) package on a finding that 
the requirements in Secs. 71.151 through 71.159 are met. The term of a 
Type B(DP) CoC is to up to 20 years.


Sec. 71.165  Conditions for package reapproval.

    (a) Except as provided in paragraph (b) of this section, each CoC 
for a Type B(DP) package or Quality Assurance Program Approval expires 
at the end of the day, in the month and year stated in the approval.
    (b) Timely renewal. If a person holding a CoC for a Type B(DP) 
package or Quality Assurance Program Approval issued under this part 
has filed a proper application requesting renewal of either the CoC or 
the Quality Assurance Program Approval, then the CoC or Quality 
Assurance Program Approval is not considered to have expired until the 
Commission has taken final action on the application. The application 
must be submitted to the Commission not less

[[Page 21458]]

than 2 years before the expiration of the CoC or the Quality Assurance 
Program Approval.
    (c) In applying for renewal of an existing CoC for a Type B(DP) 
package or Quality Assurance Program Approval, an applicant may be 
required to submit a consolidated application that incorporates all 
changes to its program--that are incorporated by reference in the 
existing approval or certificate--into as few referenceable documents 
as reasonably achievable.
    (d) Applications for renewal of an existing CoC for a Type B(DP) 
package or Quality Assurance Program Approval must be submitted to the 
Commission in accordance with Sec. 71.1.


Sec. 71.167  Application to amend a certificate of compliance.

    A certificate holder desiring to amend its CoC for a Type B(DP) 
package--including a change to the terms, conditions, or specifications 
of the CoC--shall submit an application for amendment with the 
Commission, in accordance with Sec. 71.1. The application must fully 
describe the changes desired and the reasons for these changes. The 
application should follow, as far as applicable, the form prescribed 
for an original application in Sec. 71.151.


Sec. 71.169  Issuance of an amendment to a certificate of compliance.

    In determining whether an amendment to a CoC for a Type B(DP) 
package will be issued to the applicant, the Commission will be guided 
by the considerations that govern the issuance of an initial CoC.


Sec. 71.171  Inspections and tests.

    (a) The certificate holder and applicant for a CoC for a Type B(DP) 
package shall permit, and make provisions for, the NRC to inspect the 
premises and facilities where a Type B(DP) package is designed, 
fabricated, and tested.
    (b) The certificate holder and applicant for a CoC for a Type B(DP) 
package shall make available to the NRC for inspection, upon reasonable 
notice, records kept by them pertaining to the design, fabrication, and 
testing of a Type B(DP) package.
    (c) The certificate holder and applicant for a CoC for a Type B(DP) 
package shall perform and make provisions that permit the NRC to 
perform tests that the Commission deems necessary or appropriate for 
the administration of the regulations in this part.


Sec. 71.173  Recordkeeping and reports.

    (a) Each certificate holder or applicant shall maintain any records 
and produce any reports that may be required by the conditions of the 
CoC or by the rules, regulations, and orders of the NRC in effectuating 
the purposes of the Act.
    (b) Records that are required by the regulations in this part or by 
conditions of the CoC must be maintained for the period specified by 
the appropriate regulation or the CoC conditions. If a retention period 
is not specified, the records must be maintained until the NRC 
terminates the CoC.
    (c) Any record maintained under this part may be either the 
original or a reproduced copy by any state-of-the-art method provided 
that any reproduced copy is duly authenticated by authorized personnel 
and is capable of producing a clear and legible copy after storage for 
the period specified by NRC regulations.
    (d) Each certificate holder shall maintain a record of each Type 
B(DP) package it has manufactured. The record must contain the 
following information:
    (1) The package identification number;
    (2) The package serial number;
    (3) The date fabrication of the package was commenced; and
    (4) The date fabrication of the package was completed.


Sec. 71.175  Changes.

    (a) Definitions for the purposes of this section:
    (1) Change means a modification or addition to, or removal from, a 
Type B(DP) package design or procedures that affect a design function, 
method of performing or controlling the function, or an evaluation that 
demonstrates that intended functions will be accomplished.
    (2) Departure from a method of evaluation described in the Final 
Safety Analysis Report (FSAR) (as updated) used in establishing the 
design bases or in the safety analyses means:
    (i) Changing any of the elements of the method described in the 
FSAR (as updated) unless the results of the analysis are conservative 
or essentially the same; or
    (ii) Changing from a method described in the FSAR to another method 
unless that method has been approved by NRC for the intended 
application.
    (3) A Type B(DP) package design as described in the FSAR (as 
updated) means:
    (i) The structures, systems, and components (SSC) that are 
described in the FSAR (as updated),
    (ii) The design and performance requirements for such SSCs 
described in the FSAR (as updated), and
    (iii) The evaluations or methods of evaluation included in the FSAR 
(as updated) for such SSCs which demonstrate that their intended 
function(s) will be accomplished.
    (4) Final Safety Analysis Report (as updated) means the Safety 
Analysis Report for a Type B(DP) package design as submitted, amended, 
and updated in accordance with Sec. 71.177.
    (5) Procedures as described in the FSAR (as updated) means those 
procedures that contain information described in the safety analysis 
report such as how SSCs are operated and controlled (including assumed 
operator actions and response times).
    (b) This section applies to each holder of a CoC for Type B(DP) 
package issued under this subpart.
    (c)(1) A certificate holder may make changes to a Type B(DP) 
package design, as described in the FSAR (as updated), and make changes 
in the procedures, as described in the FSAR (as updated), without 
obtaining a CoC amendment under Sec. 71.167 if:
    (i) A change in the terms, conditions, or specifications 
incorporated in the CoC is not required; and
    (ii) The change does not meet any of the criteria in paragraph 
(c)(2) of this section.
    (2) A certificate holder shall obtain a CoC amendment under 
Sec. 71.167 before implementing a proposed change, if the change would:
    (i) Result in more than a minimal increase in the frequency of 
occurrence of an accident previously evaluated in the FSAR (as 
updated);
    (ii) Result in more than a minimal increase in the likelihood of 
occurrence of a malfunction of a system, structure, or component (SSC) 
important to safety previously evaluated in the FSAR (as updated);
    (iii) Result in more than a minimal increase in the consequences of 
an accident previously evaluated in the FSAR (as updated);
    (iv) Result in more than a minimal increase in the consequences of 
a malfunction of an SSC important to safety previously evaluated in the 
FSAR (as updated);
    (v) Create a possibility for an accident of a different type than 
any previously evaluated in the FSAR (as updated);
    (vi) Create a possibility for a malfunction of an SSC important to 
safety with a different result than any previously evaluated in the 
FSAR (as updated);
    (vii) Result in a design basis limit for a fission product barrier 
as described in the FSAR (as updated) being exceeded or altered; or
    (viii) Result in a departure from a method of evaluation described 
in the FSAR (as updated) used in establishing

[[Page 21459]]

the design bases or in the safety analyses.
    (3) In implementing this paragraph, the FSAR (as updated) is 
considered to include FSAR changes resulting from evaluations performed 
under this section and analyses performed under Sec. 71.161, since the 
last update of the FSAR as required by Sec. 71.177.
    (4) The provisions in this section do not apply to changes to 
procedures when the applicable regulations of this part establish more 
specific criteria for accomplishing such changes.
    (d)(1) The certificate holder shall maintain records of changes to 
a Type B(DP) package and of changes in procedures made under paragraph 
(c) of this section. These records must include a written evaluation 
that provides the bases for the determination that the change does not 
require a CoC amendment under paragraph (c)(2) of this section.
    (2) The certificate holder shall submit, as specified in Sec. 71.1, 
a report containing a brief description of any changes, including a 
summary of the evaluation of each. A report must be submitted at 
intervals not to exceed 24 months.
    (3) The records of changes in a Type B(DP) package design must be 
maintained until:
    (i) The Commission terminates the CoC issued under this part; or
    (ii) The package is permanently removed from service.
    (4) The records of changes in procedures must be maintained for a 
period of 5 years.
    (5) The holder of a Type B(DP) package design CoC, who permanently 
ceases operation, shall provide the records of changes to the new 
certificate holder or to the Commission, in accordance with Sec. 71.1, 
as appropriate.
    (6) A certificate holder shall provide a copy of the record for any 
changes to a Type B(DP) package design to any licensee using the 
package design within 60 days of implementing the change.


Sec. 71.177  Safety analysis report updating.

    (a) Each certificate holder for a Type B(DP) package approved under 
this subpart shall update periodically, as provided in paragraph (b) of 
this section, the final safety analysis report (FSAR) to assure that 
the information included in the report contains the latest information 
developed.
    (1) Each certificate holder shall submit an original FSAR to the 
Commission, in accordance with Sec. 71.1, within 90 days after the Type 
B(DP) package design has been approved under Sec. 71.163.
    (2) The original FSAR must be based on the safety analysis report 
submitted with the application and reflect any changes and applicant 
commitments developed during the Type B(DP) package design review 
process. The original FSAR must be updated to reflect any changes to 
requirements contained in the issued CoC.
    (b) Each update must contain all the changes necessary to reflect 
information and analyses submitted to the Commission by the certificate 
holder or prepared by the certificate holder pursuant to Commission 
requirements since the submission of the original FSAR or, as 
appropriate, the last update to the FSAR under this section. The update 
must include the effects \3\ of:
---------------------------------------------------------------------------

    \3\ Effects of changes include appropriate revisions of 
descriptions in the FSAR so that the FSAR (as updated) is complete 
and accurate.
---------------------------------------------------------------------------

    (1) All changes made in the dual-purpose spent fuel transportation 
package procedures as described in the FSAR;
    (2) All safety analyses and evaluations performed by the 
certificate holder either in support of approved CoC amendments, or in 
support of conclusions that changes did not require a CoC amendment in 
accordance with Sec. 71.175; and
    (3) All analyses of new safety issues performed by or on behalf of 
the certificate holder at Commission request. The information shall be 
appropriately located within the updated FSAR.
    (c)(1) The update of the FSAR must be filed in accordance with 
Sec. 71.1, on a replacement-page basis;
    (2) The update must include a list that identifies the current 
pages of the FSAR following page replacement;
    (3) Each replacement page must include both a change indicator for 
the area changed, e.g., a bold line vertically drawn in the margin 
adjacent to the portion actually changed, and a page change 
identification (date of change or change number or both);
    (4) The update must include:
    (i) A certification by a duly authorized officer of the certificate 
holder that either the information accurately presents changes made 
since the previous submittal, or that no such changes were made; and
    (ii) An identification of changes made by the certificate holder 
under the provisions of Sec. 71.175, but not previously submitted to 
the Commission;
    (5) The update must reflect all changes implemented up to a maximum 
of 6 months before the date of filing;
    (6) Updates must be filed every 24 months from the date of issuance 
of the CoC;
    (7) Updates must be filed within 90 days of issuance from the date 
of an amendment to the CoC; and
    (8) The certificate holder shall provide a copy of the updated FSAR 
to each licensee who is using its Type B(DP) package design.
    (d) The updated FSAR must be retained by the certificate holder 
until the Commission terminates the certificate.
    (e) A certificate holder who permanently ceases operation shall 
provide the updated FSAR to the new certificate holder or to the 
Commission, in accordance with Sec. 71.1, as appropriate.
    18. Appendix A to Part 71 is revised to read as follows:

APPENDIX A TO PART 71--DETERMINATION OF A1 AND A2

    I. Values of A1 and A2 for individual 
radionuclides, which are the bases for many activity limits 
elsewhere in this chapter, are given in Table A-1. The curie (Ci) 
values specified are obtained by converting from the Terabecquerel 
(TBq) figure. The curie values are expressed to three significant 
figures to assure that the difference in the TBq and Ci quantities 
is one tenth of one percent or less. Where values of A1 
and A2 are unlimited, it is for radiation control 
purposes only. For nuclear criticality safety, some materials are 
subject to controls placed on fissile material.
    II.(a) For individual radionuclides whose identities are known, 
but which are not listed in Table A-1, the A1 and 
A2 values contained in Table A-3 may be used. Otherwise, 
the licensee shall obtain prior Commission approval of the 
A1 and A2 values for radionuclides not listed 
in Table A-1, before shipping the material.
    (b) For individual radionuclides whose identities are known, but 
which are not listed in Table A-2, the exempt material activity 
concentration and exempt consignment activity values contained in 
Table A-3 may be used. Otherwise, the licensee shall obtain prior 
Commission approval of the exempt material activity concentration 
and exempt consignment activity values, for radionuclides not listed 
in Table A-2, before shipping the material.
    (c) The licensee shall submit requests for prior approval, 
described under paragraphs II(a) and II(b) of this Appendix, to the 
Commission, in accordance with Sec. 71.1 of this part.
    III. In the calculations of A1 and A2 for 
a radionuclide not in Table A-1, a single radioactive decay chain, 
in which radionuclides are present in their naturally occurring 
proportions, and in which no daughter radionuclide has a half-life 
either longer than 10 days, or longer than that of the parent 
radionuclide, shall be considered as a single radionuclide, and the 
activity to be taken into account, and the A1 or 
A2 value to be applied shall be those corresponding to

[[Page 21460]]

the parent radionuclide of that chain. In the case of radioactive 
decay chains in which any daughter radionuclide has a half-life 
either longer than 10 days, or greater than that of the parent 
radionuclide, the parent and those daughter radionuclides shall be 
considered as mixtures of different radionuclides.
    IV. For mixtures of radionuclides whose identities and 
respective activities are known, the following conditions apply:
    (a) For special form radioactive material, the maximum quantity 
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.010

Where B(i) is the activity of radionuclide I, and A1(i) 
is the A1 value for radionuclide I.

    (b) For normal form radioactive material, the maximum quantity 
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.011

Where B(i) is the activity of radionuclide I, and A2(i) 
is the A2 value for radionuclide I.

    (c) Alternatively, the A1 value for mixtures of 
special form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.012

Where f(i) is the fraction of activity for radionuclide I in the 
mixture, and A1(i) is the appropriate A1 value 
for radionuclide I.

    (d) Alternatively, the A2 value for mixtures of 
normal form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.013

Where f(i) is the fraction of activity for radionuclide I in the 
mixture, and A2(i) is the appropriate A2 value 
for radionuclide I.

    (e) The exempt activity concentration for mixtures of nuclides 
may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.014

Where f(i) is the fraction of activity concentration of radionuclide 
I in the mixture, and [A] is the activity concentration for exempt 
material containing radionuclide I.

    (f) The activity limit for an exempt consignment for mixtures of 
radionuclides may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP30AP02.015

Where f(i) is the fraction of activity of radionuclide I in the 
mixture, and A is the activity limit for exempt consignments for 
radionuclide I.
    V. When the identity of each radionuclide is known, but the 
individual activities of some of the radionuclides are not known, 
the radionuclides may be grouped and the lowest A1 or 
A2 value, as appropriate, for the radionuclides in each 
group may be used in applying the formulas in paragraph IV. Groups 
may be based on the total alpha activity and the total beta/gamma 
activity when these are known, using the lowest A1 or 
A2 values for the alpha emitters and beta/gamma emitters.


                                                     Table A-1.--A1 and A2 Values for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                                                            Specific
    Symbol of radionuclide        Element and        A1 (TBq)           A1 (Ci)          A2 (TBq)          A2 (Ci)       activity (TBq/      Specific
                                 atomic number                                                                                 g)        activity (Ci/g)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225 (a)...................  Actinium (89)...  8.0 x 10-1......  2.2 x 10\1\.....  6.0 x 10-3......  1.6 x 10-1.....  2.1 x 10\3\....  5.8 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ac-227 (a)...................  ................  9.0 x 10-1......  2.4 x 10\1\.....  9.0 x 10-5......  2.4 x 10-3.....  2.7............  7.2 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ac-228.......................  ................  6.0 x 10-1......  1.6 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  8.4 x 10\4\....  2.2 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ag-105.......................  Silver (47).....  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  1.1 x 10\3\....  3.0 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ag-108m (a)..................  ................  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  9.7 x 10-1.....  2.6 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ag-110m (a)..................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.8 x 10\2\....  4.7 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ag-111.......................  ................  2.0.............  5.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  5.8 x 10\3\....  1.6 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Al-26........................  Aluminum (13)...  1.0 x 10-1......  2.7.............  1.0 x 10-1......  2.7............  7.0 x 10-4.....  1.9 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Am-241.......................  Americium (95)..  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  1.3 x 10-1.....  3.4
------------------------------                  --------------------------------------------------------------------------------------------------------
Am-242m (a)..................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  3.6 x 10-1.....  1.0 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Am-243 (a)...................  ................  5.0.............  1.4 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  7.4 x 10-3.....  2.0 x 10-1
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ar-37........................  Argon (18)......  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  3.7 x 10\3\....  9.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ar-39........................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  4.0 x 10\1\.....  1.1 x 10\3\....  1.3............  3.4 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ar-41........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.5 x 10\6\....  4.2 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
As-72........................  Arsenic (33)....  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  6.2 x 10\4\....  1.7 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
As-73........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  8.2 x 10\2\....  2.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
As-74........................  ................  1.0.............  2.7 x 10\1\.....  9.0 x 10-1......  2.4 x 10\1\....  3.7 x 10\3\....  9.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
As-76........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  5.8 x 10\4\....  1.6 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
As-77........................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  7.0 x 10-1......  1.9 x 10\1\....  3.9 x 10\4\....  1.0 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
At-211 (a)...................  Astatine (85)...  2.0 x 10\1\.....  5.4 x 10\2\.....  5.0 x 10-1......  1.4 x 10\1\....  7.6 x 10\4\....  2.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21461]]

 
Au-193.......................  Gold (79).......  7.0.............  1.9 x 10\2\.....  2.0.............  5.4 x 10\1\....  3.4 x 10\4\....  9.2 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Au-194.......................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  1.5 x 10\4\....  4.1 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Au-195.......................  Gold (79).......  1.0 x 10\1\.....  2.7 x 10\2\.....  6.0.............  1.6 x 10\2\....  1.4 x 10\2\....  3.7 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Au-198.......................  ................  1.0.............  2.7 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  9.0 x 10\3\....  2.4 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Au-199.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  7.7 x 10\3\....  2.1 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ba-131 (a)...................  Barium (56).....  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  3.1 x 10\3\....  8.4 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ba-133.......................  ................  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  9.4............  2.6 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ba-133m......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  2.2 x 10\4\....  6.1 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ba-140 (a)...................  ................  5.0 x 10-1......  1.4 x 10\1\.....  3.0 x 10-1......  8.1............  2.7 x 10\3\....  7.3 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Be-7.........................  Beryllium (4)...  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10\1\.....  5.4 x 10\2\....  1.3 x 10\4\....  3.5 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Be-10........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  6.0 x 10-1......  1.6 x 10\1\....  8.3 x 10-4.....  2.2 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Bi-205.......................  Bismuth (83)....  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  1.5 x 10-3.....  4.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Bi-206.......................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  3.8 x 10\3\....  1.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Bi-207.......................  ................  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  1.9............  5.2 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Bi-210.......................  ................  1.0.............  2.7 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  4.6 x 10\3\....  1.2 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Bi-210m (a)..................  ................  6.0 x 10-1......  1.6 x 10\1\.....  2.0 x 10-2......  5.4 x 10-1.....  2.1 x 10-5.....  5.7 x 10-4
------------------------------                  --------------------------------------------------------------------------------------------------------
Bi-212 (a)...................  ................  7.0 x 10-1......  1.9 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  5.4 x 10\5\....  1.5 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Bk-247.......................  Berkelium (97)..  8.0.............  2.2 x 10\2\.....  8.0 x 10-4......  2.2 x 10-2.....  3.8 x 10-2.....  1.0
------------------------------                  --------------------------------------------------------------------------------------------------------
Bk-249 (a)...................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0 x 10-1......  8.1............  6.1 x 10\1\....  1.6 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Br-76........................  Bromine (35)....  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  9.4 x 10\4\....  2.5 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Br-77........................  ................  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  2.6 x 10\4\....  7.1 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Br-82........................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  4.0 x 10\4\....  1.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
C-11.........................  Carbon (6)......  1.0.............  2.7 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  3.1 x 10\7\....  8.4 x 10\8\
------------------------------                  --------------------------------------------------------------------------------------------------------
C-14.........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0.............  8.1 x 10\1\....  1.6 x 10-1.....  4.5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ca-41........................  Calcium (20)....  1...............  1...............  1...............  1..............  3.1 x 10-3.....  8.5 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Ca-45........................  ................  4.0 x 101.......  1.1 x 103.......  1.0.............  2.7 x 101......  6.6 x 102......  1.8 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Ca-47 (a)....................  ................  3.0.............  8.1 x 101.......  3.0 x 10-1......  8.1............  2.3 x 104......  6.1 x 105
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cd-109.......................  Cadmium 48......  3.0 x 101.......  8.1 x 102.......  2.0.............  5.4 x 101......  9.6 x 101......  2.6 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Cd-113m......................  ................  4.0 x 101.......  1.1 x 103.......  5.0 x 10-1......  1.4 x 101......  8.3............  2.2 x 102
------------------------------                  --------------------------------------------------------------------------------------------------------
Cd-115 (a)...................  ................  3.0.............  8.1 x 101.......  4.0 x 10-1......  1.1 x 101......  1.9 x 104......  5.1 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Cd-115m......................  ................  5.0 x 10-1......  1.4 x 101.......  5.0 x 10-1......  1.4 x 101......  9.4 x 102......  2.5 x 104
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ce-139.......................  Cerium (58).....  7.0.............  1.9 x 102.......  2.0.............  5.4 x 101......  2.5 x 102......  6.8 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Ce-141.......................  ................  2.0 x 101.......  5.4 x 102.......  6.0 x 10-1......  1.6 x 101......  1.1 x 103......  2.8 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Ce-143.......................  ................  9.0 x 10-1......  2.4 x 101.......  6.0 x 10-1......  1.6 x 101......  2.5 x 104......  6.6 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Ce-144 (a)...................  ................  2.0 x 10-1......  5.4.............  2.0 x 10-1......  5.4............  1.2 x 102......  3.2 x 103
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21462]]

 
Cf-248.......................  Californium (98)  4.0 x 101.......  1.1 x 103.......  6.0 x 10-3......  1.6 x 10-1.....  5.8 x 101......  1.6 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-249.......................  ................  3.0.............  8.1 x 101.......  8.0 x 10-4......  2.2 x 10-2.....  1.5 x 10-1.....  4.1
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-250.......................  ................   2.0 x 101......  5.4 x 102.......  2.0 x 10-3......  5.4 x 10-2.....  4.0............  1.1 x 102
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-251.......................  ................  7.0.............  1.9 x 102.......  7.0 x 10-4......  1.9 x 10-2.....  5.9 x 10-2.....  1.6
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-252 (h)...................  ................  1.0 x 10-1......  2.7.............  1.0 x 10-3......  2.7 x 10-2.....  2.0 x 101......  5.4 x 102
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-253 (a)...................  ................  4.0 x 101.......  1.1 x 103.......  4.0 x 10-2......  1.1............  1.1 x 103......  2.9 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Cf-254.......................  ................  1.0 x 10-3......  2.7 x 10-2......  1.0 x 10-3......  2.7 x 10-2.....  3.1 x 102......  8.5 x 103
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cl-36........................  Chlorine (17)...  1.0 x 101.......  2.7 x 102.......  6.0 x 10-1......  1.6 x 101......  1.2 x 10-3.....  3.3 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Cl-38........................  ................  2.0 x 10-1......  5.4.............  2.0 x 10-1......  5.4............  4.9 x 106......  1.3 x 108
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cm-240.......................  Curium (96).....  4.0 x 101.......  1.1 x 103.......  2.0 x 10-2......  5.4 x 10-1.....  7.5 x 102......  2.0 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-241.......................  ................  2.0.............  5.4 x 101.......  1.0.............  2.7 x 101......  6.1 x 102......  1.7 x 104
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cm-242.......................  Curium (96).....  4.0 x 101.......  1.1 x 103.......  1.0 x 10-2......  2.7 x 10-1.....  1.2 x 102......  3.3 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-243.......................  ................  9.0.............  2.4 x 102.......  1.0 x 10-3......  2.7 x 10-2.....  1.9 x 10-3.....  5.2 x 101
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-244.......................  ................  2.0 x 101.......  5.4 x 102.......  2.0 x 10-3......  5.4 x 10-2.....  3.0............  8.1 x 101
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-245.......................  ................  9.0.............  2.4 x 102.......  9.0 x 10-4......  2.4 x 10-2.....  6.4 x 10-3.....  1.7 x 10-1
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-246.......................  ................  9.0.............  2.4 x 102.......  9.0 x 10-4......  2.4 x 10-2.....  1.1 x 10-2.....  3.1 x 10-1
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-247 (a)...................  ................  3.0.............  8.1 x 101.......  1.0 x 10-3......  2.7 x 10-2.....  3.4 x 10-6.....  9.3 x 10-5
------------------------------                  --------------------------------------------------------------------------------------------------------
Cm-248.......................  ................  2.0 x 10-2......  5.4 x 10-1......  3.0 x 10-4......  8.1 x 10-3.....  1.6 x 10-5.....  4.2 x 10-3
--------------------------------------------------------------------------------------------------------------------------------------------------------
Co-55........................  Cobalt (27).....  5.0 x 10-1......  1.4 x 101.......  5.0 x 10-1......  1.4 x 101......  1.1 x 105......  3.1 x 106
------------------------------                  --------------------------------------------------------------------------------------------------------
Co-56........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.1 x 103......  3.0 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Co-57........................  ................  1.0 x 101.......  2.7 x 102.......  1.0 x 101.......  2.7 x 102......  3.1 x 102......  8.4 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Co-58........................  ................  1.0.............  2.7 x 101.......  1.0.............  2.7 x 101......  1.2 x 103......  3.2 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Co-58m.......................  ................  4.0 x 101.......  1.1 x 103.......  4.0 x 101.......  1.1 x 103......  2.2 x 105......  5.9 x 106
------------------------------                  --------------------------------------------------------------------------------------------------------
Co-60........................  ................  4.0 x 10-1......  1.1 x 101.......  4.0 x 10-1......  1.1 x 101......  4.2 x 101......  1.1 x 103
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cr-51........................  Chromium (24)...  3.0 x 101.......  8.1 x 102.......  3.0 x 101.......  8.1 x 102......  3.4 x 103......  9.2 x 104
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cs-129.......................  Cesium (55).....  4.0.............  1.1 x 102.......  4.0.............  1.1 x 102......  2.8 x 104......  7.6 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-131.......................  ................  3.0 x 101.......  8.1 x 102.......  3.0 x 101.......  8.1 x 102......  3.8 x 103......  1.0 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-132.......................  ................  1.0.............  2.7 x 101.......  1.0.............  2.7 x 101......  5.7 x 103......  1.5 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-134.......................  ................  7.0 x 10-1......  1.9 x 101.......  7.0 x 10-1......  1.9 x 101......  4.8 x 101......  1.3 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-134m......................  ................  4.0 x 101.......  1.1 x 103.......  6.0 x 10-1......  1.6 x 101......  3.0 x 105......  8.0 x 106
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-135.......................  ................  4.0 x 101.......  1.1 x 103.......  1.0.............  2.7 x 101......  4.3 x 10-5.....  1.2 x 10-3
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-136.......................  ................  5.0 x 10-1......  1.4 x 101.......  5.0 x 10-1......  1.4 x 101......  2.7 x 103......  7.3 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Cs-137 (a)...................  ................  2.0.............  5.4 x 101.......  6.0 x 10-1......  1.6 x 101......  3.2............  8.7 x 101
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cu64.........................  Copper (29).....  6.0.............  1.6 x 10\2\.....  1.0.............  2.7 x 10\1\....  1.4 x 10\5\....  3.9 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Cu-67........................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  7.0 x -10-1.....  1.9 x 10\1\....  2.8 x 10\4\....  7.6 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21463]]

 
Dy-159.......................  Dysprosium (66).  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10\1\.....  5.4 x 10\2\....  2.1 x 10\2\....  5.7 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Dy-165.......................  ................  9.0 x 10-1......  2.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10-1.....  3.0 x 10\5\....  8.2 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Dy-166 (a)...................  ................  9.0 x 10-1......  2.4 x 10\1\.....  3.0 x 10-1......  8.1............  8.6 x 10\3\....  2.3 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Er-169.......................  Erbium (68).....  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0.............  2.7 x 10\1\....  3.1 x 10\3\....  8.3 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Er-171.......................  ................  8.0 x 10-1......  2.2 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  9.0 x 10\4\....  2.4 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-147.......................  Europium (63)...  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  1.4 x 10\3\....  3.7 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-148.......................  ................  5.0 x 10 -1.....  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  6.0 x 10\2\....  1.6 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-149.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10\1\.....  5.4 x 10\2\....  3.5 x 10\2\....  9.4 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-150 (Short lived).........  ................  2.0.............  5.4 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  6.1 x 10\4\....  1.6 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-150 (long lived)..........  ................  2.0.............  5.4 x 10\1\.....  7.0 x 10 -1.....  1.9 x 10\1\....  6.1 x 10\4\....  1.6 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-152.......................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  6.5............  1.8 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-152m......................  ................  8.0 x 10 -1.....  2.2 x 10\1\.....  8.0 x 10-1......  2.2 x 10\1\....  8.2 x 10\4\....  2.2 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-154.......................  ................  9.0 x -1........  2.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  9.8............  2.6 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-155.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  3.0.............  8.1 x 10\1\....  1.8 x 10\1\....  4.9 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Eu-156.......................  ................  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x -1........  1.9 x 10\1\....  2.0 x 10\3\....  5.5 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
F-18.........................  Fluorine (9)....  1.0.............  2.7 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  3.5 x 10\6\....  9.5 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Fe-52 (a)....................  Iron (26).......  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  2.7 x 10\5\....  7.3 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Fe-55........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  8.8 x 10\1\....  2.4 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Fe-59........................  ................  9.0 x 10-1......  2.4 x 10\1\.....  9.0 x -.........  2.4 x 10\1\....  1.8 x 10\3\....  5.0 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Fe-60 (a)....................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0 x 10-1......  5.4............  7.4 x 10-4.....  2.0 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ga-67........................  Gallium (31)....  7.0.............  1.9 x 10\2\.....  3.0.............  8.1 x 10\1\....  2.2 x 10\4\....  6.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ga-68........................  ................  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  1.5 x 10\6\....  4.1 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ga-72........................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.1 x 10\5\....  3.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Gd-146.......................  Gadolinium (64).  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  6.9 x 10\2\....  1.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Gd-148.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10-3......  5.4 x 10-2.....  1.2............  3.2 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Gd-153.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  9.0.............  2.4 x 10\2\....  1.3 x 10\2\....  3.5 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Gd-159.......................  ................  3.0.............  8.1 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  3.9 x 10\4\....  1.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ge-68 (a)....................  Germanium (32)..  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  2.6 x 10\2\....  7.1 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ge-71........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  5.8 x 10\3\....  1.6 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ge-77........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.3 x 10\5\....  3.6 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Hf-172 (a)...................  Hafnium (72)....  6.0 x 10-1......  1.6 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  4.1 x 10\1\....  1.1 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Hf-175.......................  ................  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  3.9 x 10\2\....  1.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Hf-181.......................  ................  2.0.............  5.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  6.3 x 10\2\....  1.7 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------

[[Page 21464]]

 
Hf-182.......................  ................  (-1)............  (-1)............  (-1)............  (-1)...........  8.1 x 10-6.....  2.2 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Hg-194 (a)...................  Mercury (80)....  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  1.3 x 10-1.....  3.5
------------------------------                  --------------------------------------------------------------------------------------------------------
Hg-195m (a)..................  ................  3.0.............  8.1 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  1.5 x 10\4\....  4.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Hg-197.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  9.2 x 10\3\....  2.5 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Hg-197m......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  4.0 x 10-1......  1.1 x 10\1\....  2.5 x 10\4\....  6.7 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Hg-203.......................  ................  5.0.............  1.4 x 10\2\.....  1.0.............  2.7 x 10\1\....  5.1 x 10\2\....  1.4 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ho-166.......................  Holmium (67)....  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  2.6 x 10\4\....  7.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ho-166m......................  ................  6.0 x 10-1......  1.6 x 10\1\.....  5.0 x 10-4......  1.4 x 10\1\....  6.6 x 10-2.....  1.8
--------------------------------------------------------------------------------------------------------------------------------------------------------
I-123........................  Iodine (53).....  6.0.............  1.6 x 10\2\.....  3.0.............  8.1 x 10\1\....  7.1 x 10\4\....  1.9 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-124........................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  9.3 x 10\3\....  2.5 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-125........................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  3.0.............  8.1 x 10\1\....  6.4 x 10\2\....  1.7 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-126........................  ................  2.0.............  5.4 x 10\1\.....  1.0.............  2.7 x 10\1\....  2.9 x 10\3\....  8.0 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-129........................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  6.5 x 10-\6\...  1.8 x 10-\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-131........................  ................  3.0.............  8.1 x 10\1\.....  7.0 x 10-\1\....  1.9 x 10\1\....  4.6 x 10\3\....  1.2 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-132........................  ................  4.0 x 10-\1\....  1.1 x 10\1\.....  4.0 x 10-\1\....  1.1 x 10\1\....  3.8 x 10\5\....  1.0 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-133........................  ................  7.0 x 10-\1\....  1.9 x 10\1\.....  6.0 x 10-\1\....  1.6 x 10\1\....  4.2 x 10\4\....  1.1 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-134........................  ................  3.0 x 10-\1\....  8.1.............  3.0 x 10-\1\....  8.1............  9.9 x 10\5\....  2.7 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
I-135 (a)....................  ................  6.0 x 10-\1\....  1.6 x 10\1\.....  6.0 x 10-\1\....  1.6 x 10\1\....  1.3 x 10\5\....  3.5 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
In-111.......................  Indium (49).....  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  1.5 x 10\4\....  4.2 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
In-113m......................  ................  4.0.............  1.1 x 10\2\.....  2.0.............  5.4 x 10\1\....  6.2 x 10\5\....  1.7 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
In-114m (a)..................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  5.0 x 10-\1\....  1.4 x 10\1\....  8.6 x 10\2\....  2.3 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
In-115m......................  ................  7.0.............  1.9 x 10\2\.....  1.0.............  2.7 x 10\1\....  2.2 x 10\5\....  6.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ir-189 (a)...................  Iridium (77)....  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  1.9 x 10\3\....  5.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ir-190.......................  ................  7.0 x 10-\1\....  1.9 x 10\1\.....  7.0 x 10-\1\....  1.9 x 10\1\....  2.3 x 10\3\....  6.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ir-192.......................  ................  1.0.............  2.7 x 10\1\.....  6.0 x 10-\1\....  1.6 x 10\1\....  3.4 x 10\2\....  9.2 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ir-194.......................  ................  3.0 x 10-\1\....  8.1.............  3.0 x 10-\1\....  8.1............  3.1 x 10\4\....  8.4 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
K-40.........................  Potassium (19)..  9.0 x 10-\1\....  2.4 x 10\1\.....  9.0 x 10-\1\....  2.4 x 10\1\....  2.4 x 10-\7\...  6.4 x 10-\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
K-42.........................  ................  2.0 x 10-\1\....  5.4.............  2.0 x 10-\1\....  5.4............  2.2 x 10\5\....  6.0 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
K-43.........................  ................  7.0 x 10-\1\....  1.9 x 10\1\.....  6.0 x 10\1\.....  1.6 x 10\1\....  1.2 x 10\5\....  3.3 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Kr-81........................  Krypton (36)....  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  7.8 x 10-\4\...  2.1 x 10-\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Kr-85........................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  1.5 x 10\1\....  3.9 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Kr-85m.......................  ................  8.0.............  2.2 x 10\2\.....  3.0.............  8.1 x 10\1\....  3.0 x 10\5\....  8.2 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Kr-87........................  ................  2.0 x 10-\1\....  5.4.............  2.0 x 10-\1\....  5.4............  1.0 x 10\6\....  2.8 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
La-137.......................  Lanthanum (57)..  3.0 x 10\1\.....  8.1 x 10\2\.....  6.0.............  1.6 x 10\2\....  1.6 x 10-\3\...  4.4 x 10-\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
La-140.......................  ................  4.0 x 10-\1\....  1.1 x 10\1\.....  4.0 x 10-\1\....  1.1 x 10\1\....  2.1 x 10\4\....  5.6 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21465]]

 
Lu-172.......................  Lutetium (71)...  6.0 x 10-\1\....  1.6 x 10\1\.....  6.0 x 10-\1\....  1.6 x 10\1\....  4.2 x 10\3\....  1.1 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Lu-173.......................  ................  8.0.............  2.2 x 10\2\.....  8.0.............  2.2 x 10\2\....  5.6 x 10\1\....  1.5 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Lu-174.......................  ................  9.0.............  2.4 x 10\2\.....  9.0.............  2.4 x 10\2\....  2.3 x 10\1\....  6.2 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Lu-174m......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  2.0 x 10\2\....  5.3 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Lu-177.......................  ................  3.0 x 10\1\.....  8.1 x 10\2\.....  7.0 x 10-\1\....  1.9 x 10\1\....  4.1 x 10\3\....  1.1 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Mg-28 (a)....................  Magnesium (12)..  3.0 x 10-\1\....  8.1.............  3.0 x 10-\1\....  8.1............  2.0 x 10\5\....  5.4 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Mn-52........................  Manganese (25)..  3.0 x 10-\1\....  8.1.............  3.0 x 10-\1\....  8.1............  1.6 x 10\4\....  4.4 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Mn-53........................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  6.8 x 10-\5\...  1.8 x 10-\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Mn-54........................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  2.9 x 10\2\....  7.7 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Mn-56........................  ................  3.0 x 10-\1\....  8.1.............  3.0 x 10-\1\....  8.1............  8.0 x 10\5\....  2.2 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Mo-93........................  Molybdenum (42).  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0 x 10\1\.....  5.4 x 10\2\....  4.1 x 10-\2\...  1.1
------------------------------                  --------------------------------------------------------------------------------------------------------
Mo-99 (a) (h)................  ................  1.0.............  2.7 x 10\1\.....  7.4 x 10-\1\....  2.0 x 10\1\....  1.8 x 10\4\....  4.8 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
N-13.........................  Nitrogen (7)....  9.0 x 10-\1\....  2.4 x 10\1\.....  6.0 x 10-\1\....  1.6 x 10\1\....  5.4 x 10\7\....  1.5 x 10\9\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Na-22........................  Sodium (11).....  5.0 x 10-\1\....  1.4 x 10\1\.....  5.0 x 10-\1\....  1.4 x 10\1\....  2.3 x 10\2\....  6.3 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Na-24........................  ................  2.0 x 10-\1\....  5.4.............  2.0 x 10-\1\....  5.4............  3.2 x 10\5\....  8.7 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Nb-93m.......................  Niobium (41)....  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0 x 10\1\.....  8.1 x 10\2\....  8.8............  2.4 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Nb-94........................  ................  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  6.9 x 10-3.....  1.9 x 10-1
------------------------------                  --------------------------------------------------------------------------------------------------------
Nb-95........................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  1.5 x 10\3\....  3.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Nb-97........................  ................  9.0 x 10-1......  2.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  9.9 x 10\5\....  2.7 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Nd-147.......................  Neodymium (60)..  6.0.............  1.6 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  3.0 x 10\3\....  8.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Nd-149.......................  ................  6.0 x 10-1......  1.6 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  4.5 x 10\5\....  1.2 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ni-59........................  Nickel (28).....  \1\.............  \1\.............  \1\.............  \1\............  3.0 x 10-3.....  8.0 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Ni-63........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0 x 10\1\.....  8.1 x 10\2\....  2.1............  5.7 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ni-65........................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  7.1 x 10\5\....  1.9 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Np-235.......................  Neptunium (93)..  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  5.2 x 10\1\....  1.4 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Np-236 (short-lived).........  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0.............  5.4 x 10\1\....  4.7 x 10-4.....  1.3 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Np-236 (long-lived)..........  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0.............  5.4 x 10\1\....  4.7 x 10-4.....  1.3 x 10-2
Np-237.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10-3......  5.4 x 10-2.....  2.6 x 10-5.....  7.1 x 10-4
------------------------------                  --------------------------------------------------------------------------------------------------------
Np-239.......................  ................  7.0.............  1.9 x 10\2\.....  4.0 x 10-1......  1.1 x 10\1\....  8.6 x 10\3\....  2.3 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Os-185.......................  Osmium (76).....  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  2.8 x 10\2\....  7.5 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Os-191.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  2.0.............  5.4 x 10\1\....  1.6 x 10\3\....  4.4 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Os-191m......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0 x 10\1\.....  8.1 x 10\2\....  4.6 x 10\4\....  1.3 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Os-193.......................  ................  2.0.............  5.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.0 x 10\4\....  5.3 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------

[[Page 21466]]

 
Os-194 (a)...................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.1 x 10\1\....  3.1 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
P-32.........................  Phosphorus (15).  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  1.1 x 10\4\....  2.9 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
P-33.........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0.............  2.7 x 10\1\....  5.8 x 10\3\....  1.6 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pa-230 (a)...................  Protactinium      2.0.............  5.4 x 10\1\.....  7.0 x 10-2......  1.9............  1.2 x 10\3\....  3.3 x 10\4\
                                (91).
------------------------------                  --------------------------------------------------------------------------------------------------------
Pa-231.......................  ................  4.0.............  1.1 x 10\2\.....  4.0 x 10-\4\....  1.1 x 10-2.....  1.7 x 10-3.....  4.7 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Pa-233.......................  ................  5.0.............  1.4 x 10\2\.....  7.0 x 10-1......  1.9 x 10\1\....  7.7 x 10\2\....  2.1 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pb-201.......................  Lead (82).......  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  6.2 x 10\4\....  1.7 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pb-202.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0 x 10\1\.....  5.4 x 10\2\....  1.2 x 10-4.....  3.4 x 10-3
------------------------------                  --------------------------------------------------------------------------------------------------------
Pb-203.......................  ................  4.0.............  1.1 x 10\2\.....  3.0.............  8.1 x 10\1\....  1.1 x 10\4\....  3.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pb-205.......................  ................  \1\.............  \1\.............  \1\.............  \1\............  4.5 x 10-6.....  1.2 x 10-4
------------------------------                  --------------------------------------------------------------------------------------------------------
Pb-210 (a)...................  ................  1.0.............  2.7 x 10\1\.....  5.0 x 10-2......  1.4............  2.8............  7.6 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pb-212 (a)...................  ................  7.0 x 10-1......  1.9 x 10\1\.....  2.0 x 10-1......  5.4............  5.1 x 10\4\....  1.4 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pd-103 (a)...................  Palladium (46)..  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  2.8 x 10\3\....  7.5 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pd-107.......................  ................  \1\.............  \1\.............  \1\.............  \1\............  1.9 x 10-5.....  5.1 x 10-4
------------------------------                  --------------------------------------------------------------------------------------------------------
Pd-109.......................  ................  2.0.............  5.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  7.9 x 10\4\....  2.1 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pm-143.......................  Promethium (61).  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  1.3 x 10\2\....  3.4 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pm-144.......................  ................  7.0 x 10-1......  1.9 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  9.2 x 10\1\....  2.5 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pm-145.......................  ................  3.0 x 10\1\.....  8.1 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  5.2............  1.4 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pm-147.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0.............  5.4 x 10\1\....  3.4 x 10\1\....  9.3 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pm-148m (a)..................  ................  8.0 x 10-1......  2.2 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  7.9 x 10\2\....  2.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pm-149.......................  ................  2.0.............  5.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  1.5 x 10\4\....  4.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pm-151.......................  ................  2.0.............  5.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.7 x 10\4\....  7.3 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Po-210.......................  Polonium (84)...  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0 x 10-2......  5.4 x 10-1.....  1.7 x 10\2\....  4.5 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pr-142.......................  Praseodymium      4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  4.3 x 10\4\....  1.2 x 10\6\
                                (59).
------------------------------                  --------------------------------------------------------------------------------------------------------
Pr-143.......................  ................  3.0.............  8.1 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.5 x 10\3\....  6.7 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pt-188 (a)...................  Platinum (78)...  1.0.............  2.7 x 101.......  8.0 x 10-1......  2.2 x 10\1\....  2.5 x 10\3\....  6.8 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-191.......................  ................  4.0.............  1.1 x 10\2\.....  3.0.............  8.1 x 10\1\....  8.7 x 10\3\....  2.4 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-193.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  1.4............  3.7 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-193m......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  5.0 x 10-1......  1.4 x 10\1\....  5.8 x 10\3\....  1.6 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-195m......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  5.0 x 10-1......  1.4 x 10\1\....  6.2 x 10\3\....  1.7 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-197.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  3.2 x 10\4\....  8.7 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pt-197m......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  3.7 x 10\5\....  1.0 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pu-236.......................  Plutonium (94)..  3.0 x 10\1\.....  8.1 x 10\2\.....  3.0 x 10-3......  8.1 x 10-2.....  2.0 x 10\1\....  5.3 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-237.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  2.0 x 10\1\.....  5.4 x 10\2\....  4.5 x 10\2\....  1.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------

[[Page 21467]]

 
Pu-238.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  6.3 x 10-1.....  1.7 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-239.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-3.....  2.3 x 10-3.....  6.2 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-240.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  8.4 x 10-3.....  2.3 x 10-1
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-241 (a)...................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  6.0 x 10-2......  1.6............  3.8............  1.0 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-242.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  1.5 x 10-4.....  3.9 x 10-3
------------------------------                  --------------------------------------------------------------------------------------------------------
Pu-244 (a)...................  ................  4.0 x 10-1......  1.1 x 10\1\.....  1.0 x 10-3......  2.7 x 10-2.....  6.7 x 10-7.....  1.8 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ra-223 (a)...................  Radium (88).....  4.0 x 10-1......  1.1 x 10\1\.....  7.0 x 10-3......  1.9 x 10-1.....  1.9 x 10\3\....  5.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ra-224 (a)...................  ................  4.0 x 10-1......  1.1 x 10\1\.....  2.0 x 10-2......  5.4 x 10-1.....  5.9 x 10\3\....  1.6 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ra-225 (a)...................  ................  2.0 x 10-1......  5.4.............  4.0 x 10-3......  1.1 x 10\1\....  1.5 x 10\3\....  3.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ra-226 (a)...................  ................  2.0 x 10-1......  5.4.............  3.0 x 10-3......  8.1 x 10-2.....  3.7 x 10-2.....  1.0
------------------------------                  --------------------------------------------------------------------------------------------------------
Ra-228 (a)...................  ................  6.0 x 10-1......  1.6 x 10\1\.....  2.0 x 10\2\.....  5.4 x 10-1.....  1.0 x 10\1\....  2.7 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rb-81........................  Rubidium (37)...  2.0.............  5.4 x 10\1\.....  8.0 x 10-1......  2.2 x 10\1\....  3.1 x 10\5\....  8.4 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rb-83 (a)....................  ................  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  6.8 x 10\2\....  1.8 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rb-84........................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  1.8 x 10\3\....  4.7 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rb-86........................  ................  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  3.0 x 10\3\....  8.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rb-87........................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  3.2 x 10-9.....  8.6 x 10-8
------------------------------                  --------------------------------------------------------------------------------------------------------
Rb(nat)......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  6.7 x 10\6\....  1.8 x 10\8\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Re-184.......................  Rhenium (75)....  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10\1\....  6.9 x 10\2\....  1.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Re-184m......................  ................  3.0.............  8.1 x 10\1\.....  1.0.............   2.7 x 10\1\...  1.6 x 10\2\....  4.3 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Re-186.......................  ................  2.0.............  5.4 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  6.9 x 10\3\....  1.9 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Re-187.......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  1.4 x 10-9.....  3.8 x 10-8
------------------------------                  --------------------------------------------------------------------------------------------------------
Re-188.......................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  3.6 x 10\4\....  9.8 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Re-189 (a)...................  ................  3.0.............  8.1 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.5 x 10\4\....  6.8 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rh-99........................  Rhodium (45)....  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  3.0 x 10\3\....  8.2 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Re(nat)......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  0.0............  2.4 x 10-8
------------------------------                  --------------------------------------------------------------------------------------------------------
Rh-101.......................  ................  4.0.............  1.1 x 10\2\.....  3.0.............  8.1 x 10\1\....  4.1 x 10\1\....  1.1 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rh-102.......................  ................  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  4.5 x 10\1\....  1.2 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rh-102m......................  ................  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  2.3 x 10\2\....  6.2 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rh-103m......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  1.2 x 10\6\....  3.3 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Rh-105.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  8.0 x 10-1......  2.2 x 1\1\.....  3.1 x 10\4\....  8.4 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rn-222 (a)...................  Radon (86)......  3.0 x 10-1......  8.1.............  4.0 x 10-3......  1.1 x 10-1.....  5.7 x 10\3\....  1.5 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
RU-97........................  Ruthenium (44)..  5.0.............  1.4 x 10\2\.....  5.0.............  1.4 x 10\2\....  1.7 x 10\4\....  4.6 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
RU-103 (a)...................  ................  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  1.2 x 10\3\....  3.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
RU-105.......................  ................  1.0.............  2.7 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.5 x 10\5\....  6.7 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
RU-106 (a)...................  ................  2.0 x 10-1......  5.4.............  2.0 x 10-1......  5.4............  1.2 x 10\2\....  3.3 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
S-35.........................  Sulphur (16)....  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0.............  8.1 x 10\1\....  1.6 x 10\3\....  4.3 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21468]]

 
Sb-122.......................  Antimony (51)...  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.5 x 10\4\....  4.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sb-124.......................  ................  6.0 x 10 -1.....  1.6 x 10\1\.....  6.0 x 10 -1.....  1.6 x 10\1\....  6.5 x 10\2\....  1.7 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sb-125.......................  ................  2.0.............  5.4 x 10\1\.....  1.0.............  2.7 x 10\1\....  3.9 x 10\1\....  1.0 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sb-126.......................  ................  4.0 x 10 -1.....  1.1 x 10\1\.....  4.0 x 10 -1.....  1.1 x 10\1\....  3.1 x 10\3\....  8.4 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sc-44........................  Scandium (21)...  5.0 x 10-1......  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  6.7 x 10\5\....  1.8 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sc-46........................  ................  5.0 x 10\1\.....  1.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  1.3 x 10\3\....  3.4 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sc-47........................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  7.0 x 10-1......  1.9 x 10\1\....  3.1 x 10\4\....  8.3 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sc-48........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  5.5 x 10\4\....  1.5 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Se-75........................  Selenium (34)...  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  5.4 x 10\2\....  1.5 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Se-79........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0.............  5.4 x 10\1\....  2.6 x 10-3.....  7.0 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Si-31........................  Silicon (14)....  6.0 x 10-1......  1.6 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  1.4 x 10\6\....  3.9 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Si-32........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  5.0 x 10-1......  1.4 x 10\1\....  3.9............  1.1 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sm-145.......................  Samarium (62)...  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10\1\.....  2.7 x 10\2\....  9.8 x 10\1\....  2.6 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sm-147.......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  8.5 x 10-1.....  2.3 x 10-8
------------------------------                  --------------------------------------------------------------------------------------------------------
Sm-151.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10\1\.....  2.7 x 10\2\....  9.7 x 10-1.....  2.6 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sm-153.......................  ................  9.0.............  2.4 x 10\2\.....  6.0 x 10-1......  1.6 x 10\1\....  1.6 x 10\4\....  4.4 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sn-113 (a)...................  Tin (50)........  4.0.............  1.1 x 10\2\.....  2.0.............  5.4 x 10\1\....  3.7 x 10\2\....  1.0 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-117m......................  ................  7.0.............  1.9 x 10\2\.....  4.0 x 10-1......  1.1 x 10\1\....  3.0 x 10\3\....  8.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-119m......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  3.0 x 10-1......  8.1 x 10\2\....  1.4 x 10\2\....  3.7 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-121m (a)..................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  9.0 x 10-1......  2.4 x 10\1\....  2.0............  5.4 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-123.......................  ................  8.0 x 10-1......  2.2 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  3.0 x 10\2\....  8.2 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-125.......................  ................  4.0 x 10-1......  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  4.0 x 10\3\....  1.1 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sn-126 (a)...................  ................  6.0 x 10-1......  1.6 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.0 x 10\3\....  2.8 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sr-82 (a)....................  Strontium (38)..  2.0 x 10-1......  5.4.............  2.0 x 10-1......  5.4............  2.3 x 10\3\....  6.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-85........................  ................  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  8.8 x 10\2\....  2.4 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-85m.......................  ................  5.0.............  1.4 x 10\2\.....  5.0.............  1.4 x 10\2\....  1.2 x 10\6\....  3.3 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-87m.......................  ................  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  4.8 x 10\5\....  1.3 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-89........................  ................  6.0 x 10-1......  1.6 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  1.1 x 10\3\....  2.9 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-90 (a)....................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  5.1............  1.4 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-91 (a)....................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.3 x 10\5\....  3.6 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Sr-92 (a)....................  ................  1.0.............  2.7 x 10\1\.....  3.0 x 10-1......  8.1............  4.7 x 10\5\....  1.3 x 10\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
T(H-3).......................  Tritium (1).....  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10-1......  1.1 x 10\3\....  3.6 x 10\2\....  9.7 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ta-178 (long-lived)..........  Tantalum (73)...  1.0.............  2.7 x 10\1\.....  8.0 x 10-1......  2.2 x 10\1\....  4.2 x 10\6\....  1.1 x 10\8\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ta-179.......................  ................  3.0 x 10-1......  8.1 x 10\2\.....  3.0 x 10\1\.....  8.1 x 10\2\....  4.1 x 10\1\....  1.1 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Ta-182.......................  ................  9.0 x 10-1......  2.4 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  2.3 x 10\2\....  6.2 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21469]]

 
Tb-157.......................  Terbium (65)....  4.0 x 10-1......  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  5.6 x 10\1\....  1.5 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tb-158.......................  ................  1.0.............  2.7 x 10\1\.....  1.0.............  2.7 x 10-1.....  5.6 x 10\1\....  1.5 x 10\1\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tb-160.......................  ................  1.0.............  2.7 x 10-1......  6.0 x 10-1......  1.6 x 10\1\....  4.2 x 10\2\....  1.1 x 10\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tc-95m (a)...................  Technetium (43).  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  8.3 x 10\2\....  2.2 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-96........................  ................  4.0 x 10\1\.....  1.1 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.2 x 10\4\....  3.2 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-96m (a)...................  ................  4.0 x 10\1\.....  1.1 x 10\1\.....  4.0 x 10\1\.....  1.1 x 10\1\....  1.4 x 10\6\....  3.8 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-97........................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  5.2 x 10\5\....  1.4 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-97m.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0.............  2.7 x 10\1\....  5.6 x 10\2\....  1.5 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-98........................  ................  8.0 x 10-1......  2.2 x 10\1\.....  7.0 x 10-1......  1.9 x 10\1\....  3.2 x 10-5.....  8.7 x 10-4
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-99........................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  9.0 x 10-1......  2.4 x 10\1\....  6.3 x 10-4.....  1.7 x 10-2
------------------------------                  --------------------------------------------------------------------------------------------------------
Tc-99m.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  4.0.............  1.1 x 10\2\....  1.9 x 10\5\....  5.3 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Te-121.......................  Tellurium (52)..  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  2.4 x 10\3\....  6.4 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-121m......................  ................  5.0.............  1.4 x 10\2\.....  3.0.............  8.1 x 10\1\....  2.6 x 10\2\....  7.0 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-123m......................  ................  8.0.............  2.2 x 10\2\.....  1.0.............  2.7 x 10\1\....  3.3 x 10\2\....  8.9 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-125m......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  9.0 x 10-1......  2.4 x 10-1.....  6.7 x 10\2\....  1.8 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-127.......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  7.0 x 10-1......  1.9 x 10\1\....  9.8 x 10\4\....  2.6 x 10\6\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-127m (a)..................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  5.0 x 10-1......  1.4 x 10\1\....  3.5 x 10\2\....  9.4 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Te-129.......................  ................  7.0 x 10\1\.....  1.9 x 10\1\.....  6.0 x 10\1\.....  1.6 x 10\1\....  7.7 x 10\5\....  2.1 x 10\7\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-129m (a)..................  ................  8.0 x 10-1......  2.2 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.1 x 10\3\....  3.0 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-131m (a)..................  ................  7.0 x 10-1......  1.9 x 10\1\.....  5.0 x 10-1......  1.4 x 10\1\....  3.0 x 10\4\....  8.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Te-132 (a)...................  ................  5.0 x 10\1\.....  1.4 x 10\1\.....  4.0 x 10-1......  1.1 x 10\1\....  1.1 x 10\4\....  8.0 x 10\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Th-227.......................  Thorium (90)....  1.0 x 10\1\.....  2.7 x 10\2\.....  5.0 x 10-3......  1.4 x 10\1\....  1.1 x 10\3\....  3.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Th-228 (a)...................  ................  5.0 x 10-1......  1.4 x 10\1\.....  1.0 x 10-3......  2.7 x 10-2.....  3.0 x 10\1\....  8.2 x 10\2\
------------------------------                  --------------------------------------------------------------------------------------------------------
Th-229.......................  ................  5.0.............  1.4 x 10\2\.....  5.0 x 10-4......  1.4 x 10-2.....  7.9 x 10-3.....  2.1 x 10-1
------------------------------                  --------------------------------------------------------------------------------------------------------
Th-230.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  1.0 x 10-3......  2.7 x 10-2.....  7.6 x 10-4.....  2.1 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Th-231.......................  Thorium (90)....  4.0 x 10\1\.....  1.1 x 10\3\.....  2.0 x 10\2\.....  5.4 x 10-1.....  2.0 x 10\4\....  5.3 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Th-232.......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  4.0 x 10-9.....  1.1 x 10-7
------------------------------                  --------------------------------------------------------------------------------------------------------
Th-234 (a)...................  ................  3.0 x 10\1\.....  8.1.............  3.0 x 10-1......  8.1............  8.6 x 10\2\....  2.3 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Th(nat)......................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  8.1 x 10-9.....  2.2 x 10-7
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ti-44 (a)....................  Titanium (22)...  5.0 x 10\1\.....  1.4 x 10\1\.....  4.0 x 10\1\.....  1.1 x 10-1.....  6.4............  1.7 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tl-200.......................  Thallium (81)...  9.0 x 10\1\.....  2.4 x 10\1\.....  9.0 x 10-1......  2.4 x 10\1\....  2.2 x 10\4\....  6.0 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tl-201.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  4.0.............  1.1 x 10\2\....  7.9 x 10\3\....  2.1 x 10\5\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tl-202.......................  ................  2.0.............  5.4 x 10\1\.....  2.0.............  5.4 x 10\1\....  2.0 x 10\3\....  5.3 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tl-204.......................  ................  1.0 x 10\1\.....  2.7 x 10\2\.....  7.0 x 10\1\.....  1.9 x 10\1\....  1.7 x 10\1\....  4.6 x 10\2\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tm-167.......................  Thulium (69)....  7.0.............  1.9 x 10\2\.....  8.0 x 10-1......  2.2 x 10\1\....  3.1 x 10\3\....  8.5 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------

[[Page 21470]]

 
Tm-170.......................  ................  3.0.............  8.1 x 10\1\.....  6.0 x 10-1......  1.6 x 10\1\....  2.2 x 10\2\....  6.0 x 10\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Tm-171.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  4.0 x 10\1\....  1.1 x 10\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
U-230 (fast lung absorption)   Uranium (92)....  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10-1......  2.7............  1.0 x 10\3\....  2.7 x 10\4\
 (a)(d).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-230 (medium lung             ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10-1......  2.7............  1.0 x 10\3\....  2.7 x 10\4\
 absorption) (a)(e).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-230 (slow lung absorption)   ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10-1......  2.7............  1.0 x 10\3\....  2.7 x 10\4\
 (a)(f).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-232 (fast lung absorption)   ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10-2......  2.7 x 10-1.....  8.3 x 10-1.....  2.2 x 10\1\
 (d).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-232 (medium lung             ................  4.0 x 10\1\.....  1.1 x 10\3\.....  1.0 x 10-2......  2.7 x 10-1.....  8.3 x 10-1.....  2.2 x 10\1\
 absorption) (e).
--------------------------------------------------------------------------------------------------------------------------------------------------------
U-233 (fast lung absorption)   Uranium (92)....  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  3.6 x 10-4.....  9.7 x 10-3
 (d).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-233 (medium lung             ................  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  3.6 x 10-4.....  9.7 x 10-3
 absorption) (e).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-233 (slow lung absorption)   ................  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  3.6 x 10-4.....  9.7 x 10-3
 (f).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-234 (fast lung absorption)   ................  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  2.3 x 10-4.....  6.2 x 10-3
 (d).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-234 (medium lung             ................  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  2.3 x 10-4.....  6.2 x 10-3
 absorption) (e).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-234 (slow lung absorption)   ................  4.0 x 101.......  1.1 x 103.......  9.0 x 10-2......  2.4............  2.3 x 10-4.....  6.2 x 10-3
 (f).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-235 (all lung absorption     ................  (1).............  (1).............  (1).............  (1)............  8.0 x 10-8.....  2.2 x 10-6
 types) (a),(d),(e),(f).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-236 (fast lung absorption)   ................  (1).............  (1).............  (1).............  (1)............  2.4 x 10-6.....  6.5 x 10-5
 (d).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-236 (medium lung             ................  (1).............  (1).............  (1).............  (1)............  2.4 x 10-6.....  6.5 x 10-5
 absorption) (e).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-236 (slow lung absorption)   ................  (1).............  (1).............  (1).............  (1)............  2.4 x 10-6.....  6.5 x 10-5
 (f).
------------------------------                  --------------------------------------------------------------------------------------------------------
U-238 (all lung absorption     ................  (1).............  (1).............  1...............  (1)............  1.2 x 10-8.....  3.4 x 10-7
 types) (d),(e),(f).
------------------------------                  --------------------------------------------------------------------------------------------------------

[[Page 21471]]

 
U (nat)......................  ................  (1).............  (1).............  (1).............  ((1))..........  2.6 x 10-8.....  7.1 x 10-7
------------------------------                  --------------------------------------------------------------------------------------------------------
U (enriched to 20% or          ................  (1).............  (1).............  (1).............  (1)............  N/A............  N/A
 less)(g).
------------------------------                  --------------------------------------------------------------------------------------------------------
U (dep)......................  ................  (1).............  (1).............  (1).............  (1)............  0.0............  (2)
--------------------------------------------------------------------------------------------------------------------------------------------------------
V-48.........................  Vanadium (23)...  4.0 x 10-1......  1.1 x 101.......  4.0 x 10-1......  1.1 x 101......  6.3 x 103......  1.7 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
V-49.........................  ................  4.0 x 101.......  1.1 x 103.......  4.0 x 101.......  1.1 x 103......  3.0 x 102......  8.1 x 103
--------------------------------------------------------------------------------------------------------------------------------------------------------
W-178 (a)....................  Tungsten (74)...  9.0.............  2.4 x 102.......  5.0.............  1.4 x 102......  1.3 x 103......  3.4 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
W-181........................  ................  3.0 x 101.......  8.1 x 102.......  3.0 x 101.......  8.1 x 102......  2.2 x 102......  6.0 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
W-185........................  ................  4.0 x 101.......  1.1 x 103.......  8.0 x 10-1......  2.2 x 101......  3.5 x 102......  9.4 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
W-187........................  ................  2.0.............  5.4 x 101.......  6.0 x 10-1......  1.6 x 101......  2.6 x 104......  7.0 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
W-188 (a)....................  ................  4.0 x 10-1......  1.1 x 101.......  3.0 x 10-1......  8.1............  3.7 x 102......  1.0 x 104
--------------------------------------------------------------------------------------------------------------------------------------------------------
 x e-122 (a).................   x enon (54)....  4.0 x 10-1......  1.1 x 101.......  4.0 x 10-1......  1.1 x 101......  4.8 x 104......  1.3 x 106
------------------------------                  --------------------------------------------------------------------------------------------------------
 x e-123.....................  ................  2.0.............  5.4 x 101.......  7.0 x 10-1......  1.9 x 101......  4.4 x 105......  1.2 x 107
------------------------------                  --------------------------------------------------------------------------------------------------------
 x e-127.....................  ................  4.0.............  1.1 x 102.......  2.0.............  5.4 x 101......  1.0 x 103......  2.8 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
 x e-131m....................  ................  4.0 x 101.......  1.1 x 103.......  4.0 x 101.......  1.1 x 103......  3.1 x 103......  8.4 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
 x e-133.....................  ................  2.0 x 101.......  5.4 x 102.......  1.0 x 101.......  2.7 x 102......  6.9 x 103......  1.9 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
 x e-135.....................  ................  3.0.............  8.1 x 101.......  2.0.............  5.4 x 101......  9.5 x 104......  2.6 x 106
--------------------------------------------------------------------------------------------------------------------------------------------------------
Y-87 (a).....................  Yttrium (39)....  1.0.............  2.7 x 101.......  1.0.............  2.7 x 101......  1.7 x 104......  4.5 x 105
--------------------------------------------------------------------------------------------------------------------------------------------------------
Y-88.........................  ................  4.0 x 10-1......  1.1 x 101.......  4.0 x 10-1......  1.1 x 101......  5.2 x 102......  1.4 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Y-90.........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  2.0 x 104......  5.4 x 105
------------------------------                  --------------------------------------------------------------------------------------------------------
Y-91.........................  ................  6.0 x 10-1......  1.6 x 101.......  6.0 x 10-1......  1.6 x 101......  9.1 x 102......  2.5 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Y-91m........................  ................  2.0.............  5.4 x 101.......  2.0.............  5.4 x 101......  1.5 x 106......  4.2 x 107
------------------------------                  --------------------------------------------------------------------------------------------------------
Y-92.........................  ................  2.0 x 10-1......  5.4.............  2.0 x 10-1......  5.4............  3.6 x 105......  9.6 x 106
------------------------------                  --------------------------------------------------------------------------------------------------------
Y-93.........................  ................  3.0 x 10-1......  8.1.............  3.0 x 10-1......  8.1............  1.2 x 105......  3.3 x 106
--------------------------------------------------------------------------------------------------------------------------------------------------------
Yb-169.......................  Ytterbium (79)..  4.0.............  1.1 x 102.......  1.0.............  2.7 x 101......  8.9 x 102......  2.4 x 104
------------------------------                  --------------------------------------------------------------------------------------------------------
Yb-175.......................  ................  3.0 x 101.......  8.1 x 102.......  9.0 x 10-1......  2.4 x 101......  6.6 x 103......  1.8 x 105
--------------------------------------------------------------------------------------------------------------------------------------------------------
Zn-65........................  Zinc (30).......  2.0.............  5.4 x 101.......  2.0.............  5.4 x 101......  3.0 x 102......  8.2 x 103
------------------------------                  --------------------------------------------------------------------------------------------------------
Zn-69........................  ................  3.0.............  8.1 x 101.......  6.0 x 10-1......  1.6 x 101......  1.8 x 106......  4.9 x 107
------------------------------                  --------------------------------------------------------------------------------------------------------
Zn-69m (a)...................  ................  3.0.............  8.1 x 101.......  6.0 x 10-1......  1.6 x 101......  1.2 x 105......  3.3 x 106
--------------------------------------------------------------------------------------------------------------------------------------------------------
Zr-88........................  Zirconium (40)..  3.0.............  8.1 x 10\1\.....  3.0.............  8.1 x 10\1\....  6.6 x 10\2\....  1.8 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Zr-93........................  ................  (\1\)...........  (\1\)...........  (\1\)...........  (\1\)..........  9.3 x 10-\5\...  2.5 x 10-\3\
------------------------------                  --------------------------------------------------------------------------------------------------------
Zr-95 (a)....................  ................  2.0.............  5.4 x 10\1\.....  8.0 x 10-\1\....  2.2 x 10\1\....  7.9 x 10\2\....  2.1 x 10\4\
------------------------------                  --------------------------------------------------------------------------------------------------------
Zr-97 (a)....................  ................  4.0 x 10-\1\....  1.1 x 10\1\.....  4.0 x 10-\1\....  1.1 x 10\1\....  7.1 x 10\4\....  1.9 x 10\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
\1\ Unlimited.
\2\ (See Table A-3.)
(a) A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days.
(b) Parent nuclides and their progeny included in secular equilibrium are listed in the following:
Sr-90 Y-90
Zr-93    Nb-93m
Zr-97    Nb-97

[[Page 21472]]

 
Ru-106    Rh-106
Cs-137    Ba-137m
Ce-134    La-134
Ce-144    Pr-144
Ba-140    La-140
Bi-212    Tl-208 (0.36), Po-212 (0.64)
Pb-210    Bi-210, Po-210
Pb-212    Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-220    Po-216
Rn-222    Po-218, Pb-214, Bi-214, Po-214
Ra-223    Rn-219, Po-215, Pb-211, Bi-211, Tl-207
Ra-224    Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Ra-226    Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
Ra-228    Ac-228
Th-226    Ra-222, Rn-218, Po-214
Th-228    Ra-224, Rn-220, Po-216, Pb212, Bi-212, Tl208 (0.36), Po-212 (0.64)
Th-229    Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
Th-nat    Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-234    Pa-234m
U-230    Th-226, Ra-222, Rn-218, Po-214
U-232    Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
U-235    Th-231
U-238    Th-234, Pa-234m
U-nat    Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214,
U-240    Np-240m
Np-237    Pa-233
Am-242m    Am-242
Am-243    Np-239
(c) The quantity may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the
  source.
(d) These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of
  transport.
(e) These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and he x avalent compounds in both normal and accident
  conditions of transport.
(f) These values apply to all compounds of uranium other than those specified in (d) and (e) above.
(g) These values apply to unirradiated uranium only.
(h) These values apply to domestic transport only. For international transport use the values in the table below.


                                                     Table A-1.--(Supplement) A1 and A2 Values for Radionuclides for International Shipments
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
                                    Element and atomic                                                                                                 Specific activity      Specific activity
      Symbol of radionuclide              number                A1 (TBq)               A1 (Ci)                A2 (TBq)               A2 (Ci)                (TBq/g)                (Ci/g)
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Cf-252...........................  Californium (98)....  5.0 x 10-2             1.4                    3.0 x 10-3             8.1 x 10-2             2.0 x 10-1             5.4 x 102
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Mo-99 (a)........................  Molybdenum (42).....  1.0                    2.7 x 101              6.0 x 10-1             1.6 x 101              1.8 x 104              4.8 x 105
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------


                      Table A-2.--Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                           Activity concentration  Activity concentration    Activity limit for      Activity limit for
      Symbol of radionuclide          Element and atomic     for exempt material     for exempt material     exempt consignment      exempt consignment
                                            number                 (Bq/g)                  (Ci/g)                   (Bq)                    (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225(a).........................  Actinium (89)........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ac-227(a).........................  .....................  1.0 x 10-1              2.7 x 10-12             1.0 x 10\3\             2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ac-228............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ag-105............................  Silver (47)..........  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ag-108m(a)........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ag-110m (a).......................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ag-111............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Al-26.............................  Aluminum (13)........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Am-241............................  Americium (95).......  1.0                     2.7 x 10-11             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Am-242m(a)........................  .....................  1.0                     2.7 x 10-11             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Am-243(a).........................  .....................  1.0                     2.7 x 10-11             1.0 x 10\3\             2.7 x 10-8
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ar-37.............................  Argon (18)...........  1.0 x 10\6\             2.7 x 10-5              1.0 x 10\8\             2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21473]]

 
Ar-39.............................  .....................  1.0 x 10\7\             2.7 x 10-4              1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ar-41.............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\9\             2.7 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
As-72.............................  Arsenic (33).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
As-73.............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
As-74.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
As-76.............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
As-77.............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
At-211(a).........................  Astatine (85)........  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\7\             2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Au-193............................  Gold (79)............  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Au-194............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Au-195............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Au-198............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Au-199............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ba-131(a).........................  Barium (56)..........  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ba-133............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ba-133m...........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ba-140(a).........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Be-7..............................  Beryllium (4)........  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Be-10.............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Bi-205............................  Bismuth (83).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bi-206............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bi-207............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bi-210............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bi-210m(a)........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bi-212(a).........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Bk-247............................  Berkelium (97).......  1.0                     2.7 x 10-11             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Bk-249(a).........................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Br-76.............................  Bromine (35).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Br-77.............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Br-82.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
C-11..............................  Carbon (6)...........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
C-14..............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\7\             2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ca-41.............................  Calcium (20).........  1.0 x 10\5\             2.7 x 10-6              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ca-45.............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21474]]

 
Ca-47(a)..........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cd-109............................  Cadmium (48).........  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cd-113m...........................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cd-115(a).........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cd-115m...........................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ce-139............................  Cerium (58)..........  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ce-141............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ce-143............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ce-144 (a)........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cf-248............................  Californium (98).....  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-249............................  .....................  1.0                     2.7 x 10-11             1.0 x 10\3\             2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-250............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-251............................  .....................  1.0                     2.7 x 10-11             1.0 x 10\3\             2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-252............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-253(a).........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cf-254............................  .....................  1.0                     2.7 x 10-11             1.0 x 10\3\             2.7 x 10-8
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cl-36.............................  Chlorine (17)........  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cl-38.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cm-240............................  Curium (96)..........  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-241............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-242............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-243............................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-244............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-245............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-246............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-247(a).........................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cm-248............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
--------------------------------------------------------------------------------------------------------------------------------------------------------
Co-55.............................  Cobalt (27)..........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Co-56.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Co-57.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Co-58.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Co-58m............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Co-60.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cr-51.............................  Chromium (24)........  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21475]]

 
Cs-129............................  Cesium (55)..........  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-131............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-132............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-134............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-134m...........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-135............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-136............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cs-137(a).........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
--------------------------------------------------------------------------------------------------------------------------------------------------------
Cu-64.............................  Copper (29)..........  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Cu-67.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Dy-159............................  Dysprosium (66)......  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Dy-165............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Dy-166(a).........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Er-169............................  Erbium (68)..........  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Er-171............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Eu-147............................  Europium (63)........  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-148............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-149............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-150 (short lived)..............  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-150 (long lived)...............  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-152............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-152m...........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-154............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-155............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Eu-156............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
F-18..............................  Fluorine (9).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Fe-52(a)..........................  Iron (26)............  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Fe-55.............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Fe-59.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Fe-60(a)..........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ga-67.............................  Gallium (31).........  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ga-68.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ga-72.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Gd-146(a).........................  Gadolinium (64)......  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21476]]

 
Gd-148............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\4\             2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Gd-153............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Gd-159............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ge-68(a)..........................  Germanium (32).......  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ge-71.............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\8\             2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ge-77.............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Hf-172(a).........................  Hafnium (72).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hf-175............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hf-181............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
Hf-182............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hg-194(a).........................  Mercury (80).........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hg-195m (a).......................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 105
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hg-197............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 7                 2.7 x 104
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hg-197m...........................  .....................  1.0 x 102               2.7 x 109               1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Hg-203............................  .....................  1.0 x 102               2.7 x 109               1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ho-166............................  Holmium (67).........  1.0 x 103               2.7 x 10-8              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ho-166m...........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
I-123.............................  Iodine (53)..........  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-124.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-125.............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-126.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               12.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-129.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-131.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-132.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-133.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
I-134.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
1I-135(a).........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
In-111............................  Indium (49)..........  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
In-113m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
In-114m(a)........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
In-115m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ir-189(a).........................  Iridium (77).........  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ir-190............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ir-192............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ir-194............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21477]]

 
K-40..............................  Potassium (19).......  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
K-42..............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
K-43..............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Kr-81.............................  Krypton (36).........  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Kr-85.............................  .....................  1.0 x 105               2.7 x 10-6              1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Kr-85m............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 1010              2.7 x 10-1
-----------------------------------                       ----------------------------------------------------------------------------------------------
Kr-87.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 109               2.7 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
La-137............................  Lanthanum (57).......  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
La-140............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Lu-172............................  Lutetium (71)          1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Lu-173............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Lu-174............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Lu-174m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Lu-177............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Mg-28(a)..........................  Magnesium (12)         1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Mn-52.............................  Manganese (25)         1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Mn-53.............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 109               2.7 x 10-2
-----------------------------------                       ----------------------------------------------------------------------------------------------
Mn-54.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Mn-56.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Mo-93.............................  Molybdenum (42)        1.0 x 103               2.7 x 10-8              1.0 x 108               2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Mo-99(a)..........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
N-13..............................  Nitrogen (7)           1.0 x 102               2.7 x 10-9              1.0 x 109               2.7 x 10-2
-----------------------------------                       ----------------------------------------------------------------------------------------------
Na-22.............................  Sodium (11)            1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Na-24.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Nb-93m............................  Niobium (41)           1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Nb-94.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Nb-95.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Nb-97.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Nd-147............................  Neodymium (60)         1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Nd-149............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ni-59.............................  Nickel (28)..........  1.0 x 104               2.7 x 10-7              1.0 x 108               2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ni-63.............................  .....................  1.0 x 105               2.7 x 10-6              1.0 x 108               2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ni-65.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Np-235............................  Neptunium (93)         1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21478]]

 
Np-236 (short-lived)..............  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Np-236 (long-lived)...............  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Np-237............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Np-239............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 102               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Os-185............................  Osmium (76)..........  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Os-191............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Os-191m...........................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Os-193............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Os-194 (a)........................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\5\             2.7 x 10-\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
P-32..............................  Phosphorus (15)......  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\5\             2.7 x 10-\6\
-----------------------------------                       ----------------------------------------------------------------------------------------------
P-33..............................  .....................  1.0 x 10\5\             2.7 x 10-\6\            1.0 x 10\8\             2.7 x 10-\3\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pa-230(a).........................  Protactinium (91)....  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pa-231............................  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pa-233............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\7\             2.7 x 10-\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pb-201............................  Lead (82)............  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pb-202............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pb-203............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pb-205............................  .....................  1.0 x 10\4\             2.7 x 10-\7\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pb-210(a).........................  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pb-212(a).........................  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\5\             2.7 x 10-\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pd-103(a).........................  Palladium (46).......  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\8\             2.7 x 10-\3\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pd-107............................  .....................  1.0 x 10\5\             2.7 x 10-\6\            1.0 x 10\8\             2.7 x 10-\3\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pd-109............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\6\             2.7 x 10-\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pm-143............................  Promethium (61)......  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-144............................  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-145............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-147............................  .....................  1.0 x 10\4\             2.7 x 10-\7\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-148m(a)........................  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-149............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pm-151............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\6\             2.7 x 10-\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Po-210............................  Polonium (84)........  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pr-142............................  Praseodymium (59)....  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\5\             2.7 x 10-\6\
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pr-143............................  .....................  1.0 x 10\4\             2.7 x 10-\7\            1.0 x 10\6\             2.7 x 10-\5\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pt-188(a).........................  Platinum (78)........  1.0 x 101               2.7 x 10-10             1.0 x 10 \6\            2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21479]]

 
Pt-191............................  .....................  1.0 x 10 \2\            2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pt-193............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pt-193m...........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pt-195m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pt-197............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pt-197m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Pu-236............................  Plutonium (94).......  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-237............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-238............................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-239............................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-240............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-241(a).........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-242............................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Pu-244(a).........................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ra-223(a).........................  Radium (88)..........  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ra-224(a).........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ra-225(a).........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ra-226(a).........................  .....................  1.0 x 101               2.7 x 10 -6             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ra-228(a).........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10 -6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rb-81.............................  Rubidium (37)........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rb-83(a)..........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rb-84.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rb-86.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rb-87.............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rb(nat)...........................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Re-184............................  Rhenium (75).........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re-184m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re-186............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re-187............................  .....................  1.0 x 106               2.7 x 10-5              1.0 x 109               2.7 x 10-2
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re-188............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10 -6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re-189(a).........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Re(nat)...........................  .....................  1.0 x 106               2.7 x 10-5              1.0 x 109               2.7 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rh-99.............................  Rhodium (45).........  1.0 x 10 \1\            2.7 x 10-10             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rh-101............................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21480]]

 
Rh-102............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rh-102m...........................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rh-103m...........................  .....................  1.0 x 10 4              2.7 x 10-7              1.0 x 10 8              2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Rh-105............................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 7              2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Rn-222(a).........................  Radon (86)...........  1.0 x 10 1              2.7 x 10-10             1.0 x 10 8              2.7 x 10-3
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ru-97.............................  Ruthenium (44).......  1.0 x 10 2              2.7 x 10-9              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ru-103(a).........................  .....................  1.0 x 10 2              2.7 x 10 -9             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ru-105............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ru-106(a).........................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 5              2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
S-35..............................  Sulphur (16).........  1.0 x 10 5              2.7 x 10-6              1.0 x 10 8              2.7 x 10-3
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sb-122............................  Antimony (51)........  1.0 x 10 2              2.7 x 10-9              1.0 x 10 4              2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sb-124............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sb-125............................  .....................  1.0 x 10 2              2.7 x 10 -9             1.0 x 10 6              2.7 x 10 -5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sb-126............................  .....................  1.0 x 10 1              2.7 x 10 -10            1.0 x 10 5              2.7 x 10 -6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sc-44.............................  Scandium (21)........  1.0 x 10 1              2.7 x 10-10             1.0 x 10 5              2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sc-46.............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sc-47.............................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sc-48.............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 5              2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Se-75.............................  Selenium (34)........  1.0 x 10 2              2.7 x 10-9              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Se-79.............................  .....................  1.0 x 10 4              2.7 x 10-7              1.0 x 10 7              2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Si-31.............................  Silicon (14).........  1.0 x 10 3              2.7 x 10-8              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Si-32.............................  .....................  1.0 x 10 3              2.7 x 10-8              1.0 x 10 6              2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sm-145............................  Samarium (62)........  1.0 x 10 2              2.7 x 10-9              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sm-147............................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 4              2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sm-151............................  .....................  1.0 x 10 4              2.7 x 10-7              1.0 x 10 8              2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sm-153............................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 6              2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sn-113(a).........................  Tin (50).............  1.0 x 10 3              2.7 x 10-8              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-117m...........................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-119m...........................  .....................  1.0 x 10 3              2.7 x 10-8              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-121m(a)........................  .....................  1.0 x 10 3              2.7 x 10-8              1.0 x 10 7              2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-123............................  .....................  1.0 x 10 3              2.7 x 10-8              1.0 x 10 6              2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-125............................  .....................  1.0 x 10 2              2.7 x 10-9              1.0 x 10 5              2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sn-126(a).........................  .....................  1.0 x 10 1              2.7 x 10-10             1.0 x 10 5              2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Sr-82(a)..........................  Strontium (38).......  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21481]]

 
Sr-85.............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-85m............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10 -4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-87m............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-89.............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-90(a)..........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-91(a)..........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Sr-92(a)..........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
T(H-3)............................  Tritium (1)..........  1.0 x 106               2.7 x 10-5              1.0 x 109               2.7 x 10-2
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ta-178 (long-lived)...............  Tantalum (73)........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ta-179............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Ta-182............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tb-157............................  Terbium (65).........  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tb-158............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tb-160............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tc-95m(a).........................  Technetium (43)......  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-96.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-96m(a).........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-97.............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 108               2.7 x 10-3
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-97m............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-98.............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-99.............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tc-99m............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Te-121............................  Tellurium (52).......  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-121m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-123m...........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-125m...........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-127............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-127m(a)........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-129............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-129m(a)........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-131m(a)........................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Te-132(a).........................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 107               2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Th-227............................  Thorium (90).........  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th-228(a).........................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------

[[Page 21482]]

 
Th-229............................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th-230............................  .....................  1.0                     2.7 x 10-11             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th-231............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 107               2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th-232............................  .....................  1.0 x 101               2.7 x 10-10             1.0 x 104               2.7 x 10-7
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th-234(a).........................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 105               2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Th(nat)...........................  .....................  1.0                     2.7 x 10-11             1.0 x 103               2.7 x 10-8
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ti-44(a)..........................  Titanium (22)........  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tl-200............................  Thallium (81)........  1.0 x 101               2.7 x 10-10             1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tl-201............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tl-202............................  .....................  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tl-204............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 104               2.7 x 10-7
--------------------------------------------------------------------------------------------------------------------------------------------------------
Tm-167............................  Thulium (69).........  1.0 x 102               2.7 x 10-9              1.0 x 106               2.7 x 10-5
Tm-170............................  .....................  1.0 x 103               2.7 x 10-8              1.0 x 106               2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Tm-171............................  .....................  1.0 x 104               2.7 x 10-7              1.0 x 108               2.7 x 10-3
--------------------------------------------------------------------------------------------------------------------------------------------------------
U-230 (fast lung absorption)        Uranium (92).........  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
 (a)(d).
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-230 (medium lung absorption)      .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
 (a)(e).
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-230 (slow lung absorption)        .....................  1.0 x 101               2.7 x 10-10             1.0 x 105               2.7 x 10-6
 (a)(f).
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-232 (fast lung absorption) (d)..  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-232 (medium lung absorption) (e)  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-232 (slow lung absorption) (f)..  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-233 (fast lung absorption) (d)..  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-233 (medium lung absorption) (e)  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-233 (slow lung absorption) (f)..  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-234 (fast lung absorption) (d)..  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-234 (medium lung absorption) (e)  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-234 (slow lung absorpiton (f)...  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-235 (all lung absorption types)   .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
 (a), (d), (e), (f).
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-236 (fast lung absorption) (d)..  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-236 (medium lung absorption (e).  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-236 (slow lung absorption) (f)..  .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U-238 (all lung absorption types)   .....................  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\4\             2.7 x 10-\7\
 (d), (e), (f).
-----------------------------------                       ----------------------------------------------------------------------------------------------
U(nat)............................  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U (enriched to 20% or less) (g)...  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
-----------------------------------                       ----------------------------------------------------------------------------------------------
U(dep)............................  .....................  1.0                     2.7 x 10-\11\           1.0 x 10\3\             2.7 x 10-\8\
--------------------------------------------------------------------------------------------------------------------------------------------------------

[[Page 21483]]

 
V-48..............................  Vanadium (23)........  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\5\             2.7 x 10-\6\
-----------------------------------                       ----------------------------------------------------------------------------------------------
V-49..............................  .....................  1.0 x 10\4\             2.7 x 10-\7\            1.0 x 10\7\             2.7 x 10-\4\
--------------------------------------------------------------------------------------------------------------------------------------------------------
W-178(a)..........................  Tungsten (74)........  1.0 x 10\1\             2.7 x 10-\10\           1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
W-181.............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
W-185.............................  .....................  1.0 x 10\4\             2.7 x 10-\7\            1.0 x 10\7\             2.7 x 10-\4\
-----------------------------------                       ----------------------------------------------------------------------------------------------
W-187.............................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\6\             2.7 x 10-\5\
-----------------------------------                       ----------------------------------------------------------------------------------------------
W-188(a)..........................  .....................  1.0 x 10\2\             2.7 x 10-\9\            1.0 x 10\5\             2.7 x 10-\6\
--------------------------------------------------------------------------------------------------------------------------------------------------------
Xe-122(a).........................  Xenon (54)...........  1.0X10\2\               2.7X10-9                1.0X10\9\               2.7X10-2
-----------------------------------                       ----------------------------------------------------------------------------------------------
Xe-123............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\9\             2.7 x 10-2
�����������������������������������
Xe-127............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\5\             2.7 x 10-6
�����������������������������������
Xe-131m...........................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\4\             2.7 x 10-7
�����������������������������������
Xe-133............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\4\             2.7 x 10-7
�����������������������������������
Xe-135............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\10\            2.7 x 10-1
--------------------------------------------------------------------------------------------------------------------------------------------------------
Y-87(a)...........................  Yttrium (39).........  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
�����������������������������������
Y-88..............................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
�����������������������������������
Y-90..............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Y-91..............................  .....................  1.0 x 10\3\             2.7 x 10-\8\            1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Y-91m.............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Y-92..............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
-----------------------------------                       ----------------------------------------------------------------------------------------------
Y-93..............................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
Yb-169............................  Ytterbium (79).......  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Yb-175............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\7\             2.7 x 10-4
--------------------------------------------------------------------------------------------------------------------------------------------------------
Zn-65.............................  Zinc (30)............  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Zn-69.............................  .....................  1.0 x 10\4\             2.7 x 10-7              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Zn-69m(a).........................  .....................  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
--------------------------------------------------------------------------------------------------------------------------------------------------------
Zr-88.............................  Zirconium (40).......  1.0 x 10\2\             2.7 x 10-9              1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Zr-93.............................  .....................  1.0 x 10\3\             2.7 x 10-8              1.0 x 10\7\             2.7 x 10-4
-----------------------------------                       ----------------------------------------------------------------------------------------------
Zr-95(a)..........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\6\             2.7 x 10-5
-----------------------------------                       ----------------------------------------------------------------------------------------------
Zr-97(a)..........................  .....................  1.0 x 10\1\             2.7 x 10-10             1.0 x 10\5\             2.7 x 10-6
--------------------------------------------------------------------------------------------------------------------------------------------------------
--------------------------------------------------------------------------------------------------------------------------------------------------------
Notes
(a) A1 and/or A2 values include contributions from daughter nuclides w/half-lives less than 10 days.
(b) Parent nuclides and their progeny included in secular equilibrium are listed in the following:
Sr-90    Y-90
Zr-93    Nb-93m
Zr-97    Nb-97
Ru-106    Rh-106
Cs-137    Ba-137m
Ce-134    La-134
Ce-144    Pr-144
Ba-140    La-140
Bi-212    Tl-208 (0.36), Po-212 (0.64)
Pb-210    Bi-210, Po-210

[[Page 21484]]

 
Pb-212    Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-220    Po-216
Rn-222    Po-218, Pb-214, Bi-214, Po-214
Ra-223    Rn-219, Po-215, Pb-211, Bi-211, Tl-207
Ra-224    Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Ra-226    Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
Ra-228    Ac-228
Th-226    Ra-222, Rn-218, Po-214
Th-228    Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36),  Po-212 (0.64)
Th-229    Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
Th-nat    Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-234    Pa-234m
U-230    Th-226, Ra-222, Rn-218, Po-214
U-232    Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
U-235    Th-231
U-238    Th-234, Pa-234m
U-nat    Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214,
U-240     Np-240m
Np-237    Pa-233
Am-242m    Am-242
Am-243    Np-239
(c) The quantity may be determined from a measurement of the rate of decay or a measurement of the radiation level at a prescribed distance from the
  source.
(d) These values apply only to compounds of uranium that take the chemical form of UF6, UO2F2 and UO2(NO3)2 in both normal and accident conditions of
  transport.
(e) These values apply only to compounds of uranium that take the chemical form of UO3, UF4, UCl4 and hexavalent compounds in both normal and accident
  conditions of transport.
(f) These values apply to all compounds of uranium other than those specified in (d) and (e) above.
(g) These values apply to unirradiated uranium only.


                                                                         Table A-3.--General Values for A \1\ and A \2\
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
                                                   A -1                                    A -2                        Activity            Activity
                                ---------------------------------------------------------------------------------  concentration for   concentration for    Activity limits     Activity limits
            Contents                                                                                                exempt material     exempt material       for exempt          for exempt
                                        (TBq)                (Ci)                (TBq)               (Ci)               (Bq/g)              (Ci/g)        consignments  (Bq)  consignments  (Ci)
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Only beta or gamma emitting      1 x 10-1             2.7 x 100           2 x 10-2            5.4 x 10-1          1 x 101             2.7 x 10-10         1 x 104             2.7 x 10-7
 radionuclides are known to be
 present.
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
Only alpha emitting              2 x 10-1             5.4 x 100           9 x 10-5            2.4 x 10-3          1 x 10-1            2.7 x 10-12         1 x 103             2.7 x 10-8
 radionuclides are known to be
 present.
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------
No relevant data are available.  1 x 10-3             2.7 x 10-2          9 x 10-5            2.4 x 10-3          1 x 10-1            2.7 x 10-12         1 x 103             2.7 x 10-8
------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------------


           Table A-4.--Activity-mass Relationships for Uranium
------------------------------------------------------------------------
                                            Specific activity
 Uranium Enrichment \1\ wt % U- ----------------------------------------
          235 present                   TBq/g                Ci/g
------------------------------------------------------------------------
0.45...........................  1.8  x  10-8         5.0  x  10-7
0.72...........................  2.6  x  10-8         7.1  x  10-7
1.0............................  2.8  x  10-8         7.6  x  10-7
1.5............................  3.7  x  10-8         1.0  x  10-6
5.0............................  1.0  x  10-7         2.7  x  10-6
10.0...........................  1.8  x  10-7         4.8  x  10-6
20.0...........................  3.7  x  10-7         1.0  x  10-5
35.0...........................  7.4  x  10-7         2.0  x  10-5
50.0...........................  9.3  x  10-7         2.5  x  10-5
90.0...........................  2.2  x  10-6         2.8  x  10-5
93.0...........................  2.6  x  10-6         7.0  x  10-5
95.0...........................  3.4  x  10-6         9.1  x  10-5
------------------------------------------------------------------------
\1\ The figures for uranium include representative values for the
  activity of the uranium-234 that is concentrated during the enrichment
  process.


    Dated at Rockville, Maryland, this 29th day of March, 2002.

    For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.

[FR Doc. 02-8108 Filed 4-29-02; 8:45 am]
BILLING CODE 7590-01-P