[Federal Register Volume 78, Number 95 (Thursday, May 16, 2013)]
[Proposed Rules]
[Pages 28988-29016]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2013-11552]



[[Page 28987]]

Vol. 78

Thursday,

No. 95

May 16, 2013

Part III





Nuclear Regulatory Commission





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10 CFR Part 71





Revisions to Transportation Safety Requirements and Harmonization With 
International Atomic Energy Agency Transportation Requirements; 
Establishing Quality Assurance Programs for Packaging Used in Transport 
of Radioactive Material; Proposed Rules

  Federal Register / Vol. 78 , No. 95 / Thursday, May 16, 2013 / 
Proposed Rules  

[[Page 28988]]


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NUCLEAR REGULATORY COMMISSION

10 CFR Part 71

[NRC-2008-0198; NRC-2013-0082]
RIN 3150-AI11


Revisions to Transportation Safety Requirements and Harmonization 
With International Atomic Energy Agency Transportation Requirements

AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule.

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SUMMARY: The U.S. Nuclear Regulatory Commission (NRC), in consultation 
with the U.S. Department of Transportation (DOT), is proposing to amend 
its regulations for the packaging and transportation of radioactive 
material. These amendments would make NRC regulations conform to 
revisions to the International Atomic Energy Agency (IAEA) regulations 
for the international transportation of radioactive material and 
maintain consistency with DOT regulations. These changes are necessary 
to maintain a consistent regulatory framework for the transportation 
and packaging of radioactive material. These changes would make the 
regulation of quality assurance programs more efficient by allowing 
changes that do not change quality assurance approval holder 
commitments to be made without prior NRC approval, and extending the 
duration of quality assurance program approvals. These changes would 
clarify the responsibilities of general licensees and further limit the 
shipping of fissile material under a general license. The parallel DOT 
proposed rulemaking was published in the Federal Register on August 12, 
2011.

DATES: Submit comments by July 30, 2013. Submit comments specific to 
the information collections aspect of this proposed rule by June 17, 
2013. Comments received after these dates will be considered if it is 
practical to do so, but the NRC is able to assure consideration only 
for comments received on or before these dates.

ADDRESSES: You may submit comments by any of the following methods 
(unless this document describes a different method for submitting 
comments on a specific topic):
     Federal rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2008-0198. Address 
questions about NRC dockets to Carol Gallagher; telephone: 301-492-
3668; email: [email protected]. For technical questions, contact 
the individual listed in the FOR FURTHER INFORMATION CONTACT section of 
this document.
     Email comments to: [email protected]. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     Fax comments to: Secretary, U.S. Nuclear Regulatory 
Commission at 301-415-1101.
     Mail comments to: Secretary, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, ATTN: Rulemakings and 
Adjudications Staff.
     Hand deliver comments to: 11555 Rockville Pike, Rockville, 
Maryland 20852, between 7:30 a.m. and 4:15 p.m. (Eastern Time) Federal 
workdays; telephone: 301-415-1677.
    For additional direction on accessing information and submitting 
comments, see ``Accessing Information and Submitting Comments'' in the 
SUPPLEMENTARY INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: James Firth, Office of Federal and 
State Materials and Environmental Management Programs, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, telephone: 301-415-
6628; email: [email protected].

SUPPLEMENTARY INFORMATION: The parallel DOT proposed rulemaking was 
published in the Federal Register on August 12, 2011 (76 FR 50332).
I. Accessing Information and Submitting Comments
II. Background
III. Discussion
    A. What action is the NRC proposing to take?
    B. Who is affected by this proposed action?
    C. Which changes are being proposed to increase the 
compatibility with the International Atomic Energy Agency 
Regulations (TS-R-1) and consistency with the DOT regulations?
    D. How is the NRC proposing to change the exemption for 
materials with low activity levels?
    E. How might the qualification of special form radioactive 
material change?
    F. What changes may be made to Appendix A, ``Determination of 
A1 and A2 Values,'' to part 71 of title 10 of 
the Code of Federal Regulations (10 CFR)?
    G. How would the responsibilities of certificate holders and 
licensees change with these amendments?
    H. Why would renewal of my quality assurance program description 
not be necessary?
    I. What changes could be made to a quality assurance program 
description without seeking prior NRC approval?
    J. How frequently would I submit periodic updates on my quality 
assurance program description to the NRC?
    K. How would the requirements in subpart H, ``Quality 
Assurance,'' change with the removal of the footnote in 10 CFR 
71.103?
    L. What changes would be made to general licenses?
    M. How would the exemption from classification as fissile 
material (10 CFR 71.15) change?
    N. What other changes is the NRC proposing to make to its 
regulations for the packaging and transportation of radioactive 
material?
    O. When Would these proposed amendments become effective?
    P. What should I consider as I prepare my comments to the NRC?
IV. Section-by-Section Analysis
V. Criminal Penalties
VI. Agreement State Compatibility
VII. Availability of Documents
VIII. Plain Writing
IX. Voluntary Consensus Standards
X. Finding of No Significant Environmental Impact: Availability
XI. Paperwork Reduction Act Statement
XII. Regulatory Analysis
XIII. Regulatory Flexibility Certification
XIV. Backfitting

I. Accessing Information and Submitting Comments

A. Accessing Information

    Please refer to Docket ID NRC-2008-0198 when contacting the NRC 
about the availability of information for this proposed rule. You may 
access information related to this proposed rulemaking, which the NRC 
possesses and is publicly available, by any of the following methods:
     Federal Rulemaking Web site: Go to http://www.regulations.gov and search for Docket ID NRC-2008-0198.
     NRC Agencywide Documents Access and Management System 
(ADAMS): You may access publicly available documents online in the NRC 
Library at http://www.nrc.gov/reading-rm/adams.html. To begin the 
search, select ``ADAMS Public Documents'' and then select ``Begin Web-
based ADAMS Search.'' For problems with ADAMS, please contact the NRC 
Public Document Room (PDR) reference staff at 1-800-397-4209, 301-415-
4737, or by email to [email protected]. The ADAMS accession number 
for each document referenced in this proposed rule (if that document is 
available in ADAMS) is provided the first time that a document is 
referenced. In addition, for the convenience of the reader, the ADAMS 
accession numbers are provided in a table in Section VI, Availability 
of Documents, of this document.
     NRC PDR: You may examine and purchase copies of public 
documents at the NRC PDR, Room O1-F21, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland 20852.

[[Page 28989]]

B. Submitting Comments

    Please include Docket ID NRC-2008-0198 in the subject line of your 
comment submission, in order to ensure that the NRC is able to make 
your comment submission available to the public in this docket.
    The NRC cautions you not to include identifying or contact 
information that you do not want publicly disclosed in your comment 
submission. The NRC will post all comment submissions at http://www.regulations.gov as well as enter the comment submissions into 
ADAMS, and the NRC does not routinely edit comment submissions to 
remove identifying or contact information.
    If you are requesting or aggregating comments from other persons 
for submission to the NRC, then you should inform those persons not to 
include identifying or contact information that they do not want to be 
publicly disclosed in their comment submission. Your request should 
state that the NRC does not routinely edit comment submissions to 
remove such information before making the comment submissions available 
to the public or entering the comment submissions into ADAMS.

II. Background

    The NRC is proposing to revise its regulations for the safe 
transportation of radioactive material to make them compatible with 
those of the IAEA. The proposed rule, in combination with a 
corresponding amendment of Title 49 of the Code of Federal Regulations 
(49 CFR), by the DOT (76 FR 50332; August 12, 2011), would bring United 
States regulations into general accord with the 2009 edition of the 
IAEA's ``Regulations for the Safe Transport of Radioactive Material'' 
(TS-R-1). The NRC is also proposing to make revisions to maintain 
consistency with revisions to DOT regulations. In addition, the NRC is 
making other revisions to its transportation regulations in 10 CFR part 
71. These other revisions include NRC-initiated changes that would 
affect administrative procedures for the quality assurance program 
requirements described in 10 CFR part 71, subpart H; re-establish 
restrictions on material that qualifies for the fissile material 
exemption; clarify the requirements for a general license; clarify the 
responsibilities of certificate holders and licensees when making 
preliminary determinations; and make other editorial changes.

Compatibility With IAEA and Consistency With DOT Transportation 
Regulations

    The IAEA was formed by member nations to promote safe, secure, and 
peaceful nuclear technologies. It establishes safety standards to 
protect public health and safety and to minimize the danger to life and 
property. The IAEA has developed international safety standards for the 
safe transport of radioactive material, TS-R-1. The IAEA safety 
standards and regulations are developed in consultation with the 
competent authorities of Member States, so they reflect an 
international consensus on what is needed to provide for a high-level 
of safety. By providing a global framework for the consistent 
regulation of the transport of radioactive material, TS-R-1 facilitates 
international commerce and contributes to the safe conduct of 
international trade involving that material. By periodically revising 
its regulations to be compatible with IAEA and DOT regulations, the NRC 
is able to remove inconsistencies that could impede international 
commerce and reflect knowledge gained in scientific and technical 
advances and accumulated experience.
    On January 26, 2004 (69 FR 3698), the NRC published in the Federal 
Register a final revision to 10 CFR part 71, ``Compatibility with IAEA 
Transportation Safety Standards (TS-R-1) and Other Transportation 
Safety Amendments.'' That revision, in combination with a parallel 
revision of the DOT hazardous materials transportation regulations, 
brought the United States domestic transport regulations into general 
accord with the 1996 edition of TS-R-1 (as amended in 2000). The DOT 
published its corresponding revision to 49 CFR parts 171 through 178 on 
the same date (69 FR 3632; January 26, 2004).
    The IAEA periodically reviews and revises the IAEA international 
transportation standards to reflect knowledge gained in scientific and 
technical advances and accumulated experience. In 2002, the IAEA began 
using a 2-year review cycle. In each review cycle, the IAEA will invite 
Member States--the United States is a Member State and the DOT is the 
United States competent authority before the IAEA for radioactive 
material transportation matters--to submit for consideration issues or 
problems that could result in changes to the IAEA transportation 
regulations and the associated guidance. These issues and problems are 
then considered by the IAEA Transportation Safety Standards Committee 
(TRANSSC) and, if approved by TRANSSC, will be developed into specific 
proposed changes to the transportation regulations. The specific 
proposed changes are then considered at a second TRANSSC meeting. The 
IAEA will then issue those approved changes at the second TRANSSC 
meeting for formal review and comment by Member States.
    The IAEA has invited Member States to submit comments and suggest 
changes to the regulations as part of these periodic revisions. The NRC 
and DOT have sought public input related to the proposed revisions. On 
July 22, 2003, the DOT held a public meeting, with the NRC 
participating, to obtain public views on proposed changes to the 1996 
edition of TS-R-1 and accepted written comments through August 8, 2003. 
On November 5, 2003, the DOT held a public meeting, with the NRC 
participating, seeking public views on the DOT positions on the 
proposed changes to TS-R-1. The NRC published Federal Register notices 
on June 26, 2003 (68 FR 37986); October 24, 2003 (68 FR 60886); April 
23, 2004 (69 FR 21978); April 27, 2005 (70 FR 21684); and November 21, 
2007 (72 FR 65470), soliciting public input on proposed revisions to 
TS-R-1. Subsequent to the 1996 edition of TS-R-1 (as amended in 2000), 
the IAEA published revisions to TS-R-1 in 2003, 2005, and 2009.
    This rulemaking effort would involve harmonizing the NRC 
regulations at 10 CFR part 71 with changes to the IAEA transportation 
regulations through TS-R-1. Copies of TS-R-1 may be obtained from the 
United States distributors, Bernan, 15200 NBN Way, P.O. Box 191, Blue 
Ridge Summit, PA 17214; telephone: 1-800-865-3457; email: 
[email protected], or Renouf Publishing Company Ltd., 812 Proctor 
Ave., Ogdensburg, NY 13669-2205; telephone: 1-888-551-7470; email: 
[email protected]. An electronic copy may be found at the 
following IAEA Web site: http://www-pub.iaea.org/MTCD/publications/PDF/Pub1384_web.pdf. The regulations in TS-R-1 represent an accepted set 
of requirements that provide a high level of safety in the packaging 
and transportation of radioactive materials and provide a basis and 
framework that facilitates the development of internationally-
consistent regulations. Internationally-consistent regulations for the 
transportation and packaging of radioactive material reduce impediments 
to trade; facilitate international cooperation; and, when the 
regulations provide a high level of safety, can reduce risks associated 
with the import and export of radioactive

[[Page 28990]]

material. Harmonization represents the effort to increase the 
consistency or compatibility between national regulations and the 
internationally-accepted requirements, within the constraints of an 
existing national legal and regulatory framework.
    In November 2012, the IAEA issued new standards for the safe 
transport of radioactive material and designated them as ``Specific 
Safety Requirements Number SSR-6'' (SSR-6). This proposed rulemaking 
would not incorporate the 2012 changes, which will undergo a 
comprehensive review by the NRC staff to determine if additional 
changes to 10 CFR part 71 are warranted.
    Historically, the NRC has coordinated its revisions to 10 CFR part 
71 with the DOT, because the DOT is the United States competent 
authority for transportation of hazardous materials. ``Radioactive 
Materials'' is a subset of ``Hazardous Materials'' in Title 49 
regulations under DOT authority. The DOT hazardous materials 
regulations are found in 49 CFR parts 171 through 177. Currently, the 
DOT and the NRC co-regulate transport of radioactive materials in the 
United States. The roles of the DOT and the NRC in the co-regulation of 
the transportation of radioactive materials are described in a 
memorandum of understanding (MOU) (44 FR 38690; July 2, 1979). 
Consistent with this MOU, the NRC is continuing to coordinate its 
efforts with the DOT in this proposed rulemaking process. Refer to the 
DOT corresponding rule for additional background on the proposed 
changes in this document.

Scope of 10 CFR Part 71 Proposed Rulemaking

    The NRC staff evaluated recent changes in the IAEA's transportation 
standards through the 2009 edition of TS-R-1 to identify changes to be 
made in 10 CFR part 71. Based on this effort, the NRC staff identified 
a number of areas in 10 CFR part 71 that need to be addressed in this 
proposed rulemaking process as a result of the changes to the IAEA 
regulations. These changes are discussed in Section III of this 
document, question C, ``Which Changes are Being Proposed to Increase 
the Compatibility with the International Atomic Energy Agency 
Regulations (TS-R-1) and Consistency with DOT Regulations?''
    The NRC is also proposing a number of self-initiated changes to its 
regulations that are not related to either compatibility with IAEA 
regulations or consistency with DOT regulations. These NRC changes 
would affect administrative procedures for the quality assurance 
program requirements described in 10 CFR part 71, subpart H, re-
establish restrictions on material that qualifies for the fissile 
material exemption, clarify the requirements for a general license, 
clarify the responsibilities of certificate holders and licensees when 
making preliminary determinations, and make other editorial changes.

Fissile Material Exemption

    In 1997, the NRC issued an emergency final rule (62 FR 5907; 
February 10, 1997) that revised the regulations on fissile material 
exemptions and the general licenses that apply to fissile material. The 
NRC determined that good cause existed under Section 553(b)(3)(B) of 
the Administrative Procedure Act (APA) (5 U.S.C. 553(b)(3)(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 the 
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 final rule. Based on the public comments on the emergency 
final rule, the 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,\1\ and the 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 concluded that the 
regulations may be excessive for shipments where water moderation is 
the only concern. The ORNL study recommended that the NRC revise 10 CFR 
part 71. The ORNL made a recommendation that applied to the requirement 
specific to uranium enriched in uranium-235 (U-235) to a maximum of 1 
percent by weight, and with a total plutonium and uranium-233 (U-233) 
content of up to 1 percent of the mass of U-235, hereafter referred to 
as uranium enriched to a maximum of 1 percent. Specifically, ORNL 
recommended: (1) That a definition of ``homogeneity'' be developed that 
could be clearly understood for use with uranium enriched to a maximum 
of 1 percent; (2) the term ``lattice arrangement'' be clarified or not 
used; and (3) if the definitions for homogeneity and lattice 
arrangement cannot be provided, a restriction on beryllium (Be), 
deuterium oxide (e.g., D2O or heavy water), and carbon 
(graphite) (C) should be maintained. The ORNL recommended that the 
moderator criteria restricting the mass of Be, C, or D2O to 
less than 0.1 percent of the fissile mass should be maintained, which 
would remove the need to provide definitions--such as ``homogeneous'' 
and ``lattice arrangement''--that are difficult to define and to apply 
practically. The NRC staff indicated that it agreed with the ORNL 
recommendations (67 FR 21390; April 30, 2002) and removed the 
homogeneity and lattice prevention requirements from the fissile 
material exemptions.
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    \1\ NUREG/CR-5342, ``Assessment and Recommendations for Fissile-
Material Packaging Exemptions and General Licenses within 10 CFR 
Part 71,'' July 1998.
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    The ORNL recommendations were considered when the NRC proposed 
changes to 10 CFR part 71 (67 FR 21390; April 30, 2002) to make NRC 
regulations more consistent and compatible with IAEA regulations and to 
make changes to the fissile material exemption requirements to address 
the unintended economic impact of the NRC emergency final rule entitled 
``Fissile Material Shipments and Exemptions'' (62 FR 5907; February 10, 
1997). In its final rule (69 FR 3698; January 26, 2004) to make 10 CFR 
part 71 compatible with the IAEA regulations and make other 
transportation safety amendments, the NRC removed the restriction that, 
to qualify for the fissile material exemption, uranium enriched in U-
235 is distributed homogeneously throughout the package and does not 
form a lattice arrangement within the package, and redesignated the 
section for fissile material exemptions from Sec.  71.53 to Sec.  
71.15. Based on a comment that shippers would have difficulty 
implementing the proposed rule language, the NRC determined that it 
would be impractical to implement a restriction based on the proposed 
ratio of the restricted moderators to the fissile mass and changed the 
restriction to require that the mass of beryllium, graphite, and 
hydrogenous material

[[Page 28991]]

enriched in deuterium be less than 5 percent of the mass of uranium; 
the NRC concluded that limiting the mass of these moderators to less 
than 5 percent of the uranium mass would assure subcriticality for all 
moderators of concern.
    Subsequent to the 2004 rulemaking, the U.S. Department of Energy 
(DOE) was planning a shipment of large quantities of low-enriched 
fissile material that would qualify for the exemption at 10 CFR 
71.15(d). Analyses performed by the DOE indicated that large arrays of 
heterogeneous uranium with enrichment of 1 percent by weight of U-235 
could exceed a keff of 0.95 when optimally moderated by 
water. For the material to become critical,\2\ the keff 
would need to be greater than or equal to 1.0. However, the quantity 
and geometric arrangement of this material exceeded a keff 
of 0.95, which is typically used as a limit in regulatory assessments 
of package designs for the transport of fissile material. The 
sensitivity of keff to increases in the quantity of fissile 
material and changes in geometry will depend on the properties of the 
material. For uranium enriched to a maximum of 1 percent and 
keff greater than 0.95, keff is very insensitive 
to changes in geometry and quantity; consequently, significantly larger 
quantities of material would be required to get keff close 
to 1.0.
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    \2\ 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. The neutron moderating material cannot 
achieve criticality by itself in any concentration or configuration. 
It can enhance the ability of fissile material to achieve 
criticality by slowing down neutrons or reflecting neutrons.
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Quality Assurance Program Approvals

    Part 71 of 10 CFR does not include provisions for making changes to 
an approved quality assurance program without obtaining prior NRC 
approval before implementing the change. The requirement to obtain 
prior NRC approval currently applies to all changes, no matter how 
insignificant in importance they are to safety. Consequently, the 
process can be overly burdensome and inefficient for both the licensee 
and the NRC. For example, a change in the quality assurance program to 
correct typographical errors or punctuation would need to be submitted 
and approved by the NRC.
    In the past, the NRC observed several instances in which holders of 
a 10 CFR part 71 quality assurance program approval had made changes to 
their NRC-approved quality assurance program before obtaining NRC 
approval. Although many of the changes were found acceptable by the NRC 
after they were reviewed, some of the changes did not satisfy the 
respective requirements of 10 CFR part 71, subpart H. In Information 
Notice 2002-35 (December 20, 2002; ADAMS Accession No. ML023520339), 
the NRC indicated that it was considering changes to 10 CFR part 71 to 
provide a method similar to 10 CFR 50.54(a)(3) and (4) for making 
changes to 10 CFR part 71 quality assurance programs.
    In 2004, the NRC changed the renewal period for quality assurance 
program approvals issued under 10 CFR part 71 from 5 years to 10 years. 
This change was announced in ``NRC Regulatory Information Summary (RIS) 
2004-18, Expiration Date for 10 CFR Part 71 Quality Assurance Program 
Approvals'' (December 1, 2004; ADAMS Accession No. ML042160293). After 
making this change, the NRC evaluated whether a change should be made 
in the regulations to codify the effective term of the quality 
assurance program approval or whether any expiration date for the 
quality assurance program approval was necessary.
    In the proposed rule section of this issue of the Federal Register, 
the NRC is issuing for public comment Draft Regulatory Guidance (DG) 
7009, ``Establishing Quality Assurance Programs for Packaging Used in 
Transport of Radioactive Material'' (RIN 3150-AI11; NRC-2013-0082).

III. Discussion

A. What action is the NRC proposing to take?

    The NRC is proposing to amend its regulations to make them more 
consistent or compatible with the IAEA international transportation 
regulations. These changes are in response to changes introduced in the 
1996 (as amended in 2003), 2005, and 2009 editions of TS-R-1. The NRC 
is proposing to revise its regulations to be consistent with DOT 
hazardous materials regulations to maintain a consistent framework for 
the transportation and packaging of radioactive material.
    The NRC is proposing to make changes that would clarify the 
requirements to obtain a general license and the responsibilities of 
general licensees. The NRC is proposing to make changes that would 
clarify the roles of users of NRC-approved packaging and certificate 
holders or applicants for a certificate of compliance (CoC). Also, the 
NRC is proposing to make changes that would make the regulation of 
quality assurance programs more efficient. The NRC is proposing to 
issue quality assurance program approvals that would not expire, 
removing the need for the approval to be renewed, and would revise the 
current quality assurance program approvals so that they would not 
expire. The NRC is also proposing to allow those changes that do not 
reduce the commitments in an approved quality assurance program to be 
made without prior NRC approval.
    The NRC is proposing to make changes that would change the 
responsibilities of licensees and certificate holders for making the 
preliminary determinations in Sec.  71.85.
    Other proposed changes would correct errors and clarify the 
regulations.

B. Who is affected by this proposed action?

    This action would affect NRC licensees authorized by a 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, or transports the material outside of the site 
of usage as specified in the NRC license, or transports that material 
on public highways; holders of, and applicants for, a CoC; and holders 
of a 10 CFR part 71, Subpart H, quality assurance program approval. 
This action would also affect holders of quality assurance program 
approvals under appendix B of 10 CFR part 50 or subpart G of 10 CFR 
part 72 to the extent that those approvals apply to transport packaging 
as specified in 10 CFR 71.101(f), ``Previously approved programs.'' 
This action would change requirements that are matters of 
compatibility. Agreement States would be required to update their 
regulations and Agreement State licensees would be affected by the 
changes to the Agreement State regulations.

C. Which changes are being made to increase the compatibility with the 
International Atomic Energy Agency Regulations (TS-R-1) and consistency 
with DOT regulations?

    The NRC has identified changes in 10 CFR part 71 that would make 
the NRC regulations more consistent or compatible with the 
international transportation regulations. These changes would also 
improve the consistency with the current DOT regulations or would 
maintain consistency between 10 CFR part 71 and DOT regulations by 
making changes that correspond to those proposed by the

[[Page 28992]]

DOT. The NRC is proposing the following changes to 10 CFR part 71.
    1. In the 2003 Edition of TS-R-1, the IAEA changed the scope of TS-
R-1 as it applies to natural materials and ores by adding language that 
addresses the processing of these materials (paragraph 107(e) of the 
2009 edition of TS-R-1). The NRC is proposing to include the concept of 
processing into the provisions that apply to natural materials and ores 
in the exemptions for low-level materials at Sec.  71.14.
    2. The NRC is proposing to adopt the scoping statement in paragraph 
107(f) of TS-R-1, which addresses non-radioactive solid objects with 
radioactive substances present on any surface in quantities not in 
excess of certain levels. In conjunction with this proposed change, a 
definition of ``contamination'' corresponding to the definition in TS-
R-1 would be added to Sec.  71.4.
    3. The NRC is proposing to amend the following definitions in 10 
CFR 71.4 to reflect the current definitions in TS-R-1: ``Criticality 
Safety Index (CSI);'' ``Low Specific Activity (LSA) material;'' and 
``uranium--natural, depleted, enriched.'' When the NRC last revised the 
definition for LSA material, the NRC added the modifier ``not,'' which 
resulted in the NRC definition becoming inconsistent with the DOT and 
IAEA definitions. The NRC is proposing to correct this, so that LSA 
material includes material intended to be processed for its 
radionuclides.
    4. The NRC is proposing to adopt the use of the Class 5 impact test 
prescribed in the International Organization for Standardization (ISO) 
document 2919, ``Radiation protection--Sealed radioactive sources--
General requirements and classification,'' Second Edition (February 15, 
1999), ISO 2919:1999(E), for special form radioactive material, 
provided the mass was less than 500 grams.
    5. The NRC is proposing to incorporate by reference ISO document 
2919, ``Radiation protection--Sealed radioactive sources--General 
requirements and classification,'' Second Edition (February 15, 1999), 
ISO 2919:1999(E), and ISO document 9978, ``Radiation protection--Sealed 
radioactive sources--Leakage test methods,'' First Edition (February 
15, 1992), ISO 9978:1992(E).
    6. The NRC is proposing to change the description of billet used in 
the percussion test in Sec.  71.75(b)(2)(ii) by replacing ``edges'' 
with ``edge.''
    7. The NRC is proposing to revise the definition of ``special form 
radioactive material'' in Sec.  71.4 to allow special form radioactive 
material that is successfully tested in accordance with the current 
requirements to continue to be transported as special form radioactive 
material, if the testing was completed before the effective date of the 
final rule.
    8. In appendix A, Table A-1, the NRC is proposing to eliminate the 
A1 and A2 values for californium-252 (Cf-252) for 
domestic use. The A1 and A2 values for Cf-252 
would be consistent with the IAEA values.
    9. The NRC is proposing to include krypton-79 (Kr-79) in Table A-1 
and Table A-2. The A1 and A2 values in Table A-1 
and the activity concentration for exempt material and the activity 
limit for exempt consignment would be consistent with the IAEA values 
in the 2009 edition of TS-R-1.
    10. The NRC is proposing to revise footnote a to Table A-1, 
``A1 and A2 values for radionuclides,'' to 
include the list of parent radionuclides whose A1 and 
A2 values include contributions from daughter radionuclides 
with half-lives of less than 10 days in footnote a to Table 2, ``Basic 
Radionuclide Values,'' in TS-R-1 (2009 edition), with the exception of 
argon-42 (Ar-42) and tellurium-118 (Te-118), which appear in footnote a 
to Table 2 in TS-R-1 (2009 edition), but do not appear within Table 2.
    11. The NRC is proposing to move and revise footnote c to Table A-1 
to make clear that only for iridium-192 (Ir-192) in special form is it 
appropriate for the activity of Ir-192 to be determined from a 
measurement of the rate of decay or a measurement of the radiation 
level at a prescribed distance.
    12. In appendix A, Table A-2, the NRC is proposing to revise the 
activity limit for exempt consignment for tellurium-121m (Te-121m) to 
be consistent with the new IAEA value.
    13. The NRC is proposing to revise the list of parent radionuclides 
and their progeny included in secular equilibrium in footnote b to 
Table A-2, ``Exempt material activity concentrations and exempt 
consignment activity limits for radionuclides,'' to be consistent with 
the list accompanying Table 2, ``Basic Radionuclide Values,'' in TS-R-1 
(2009 edition).
    14. The NRC is proposing to revise the descriptive phrases for 
different categories of unknown radionuclides and mixtures in Table A-3 
to be consistent with the IAEA descriptions in Table 3, ``Basic 
Radionuclide Values for Unknown Radionuclides or Mixtures,'' in TS-R-1 
(2009 edition). The descriptive phrases for ``Only alpha emitting 
nuclides are known to be present'' and ``No relevant data are 
available'' would be revised.

D. How is the NRC proposing to change the exemption for materials with 
low activity levels?

    The NRC is proposing to revise its exemption for natural materials 
and ores containing naturally occurring radionuclides to reflect 
changes in the scope of TS-R-1. In its proposed rule (76 FR 50332; 
August 12, 2011), the DOT proposed adopting these changes.
    The TS-R-1 includes statements that describe its scope. First, 
there is a description of activities included within the scope of 
regulation. Second, TS-R-1 has a list of material to which TS-R-1 does 
not apply, hereafter referred to as ``non-TS-R-1 material.'' Included 
in the list of non-TS-R-1 material are natural materials and ores 
containing naturally occurring radionuclides. These natural materials 
and ores are not intended to be processed for their radionuclides, 
provided that the activity concentration for the material does not 
exceed 10 times the activity concentration for exempt material. In the 
2003 edition of TS-R-1, the description of natural materials and ores 
containing naturally occurring radionuclides contained in the list of 
non-TS-R-1 material was revised to add natural materials and ores that 
have been processed.
    In the 2003 edition of TS-R-1, ``non-radioactive solid objects with 
radioactive substances on any surfaces'' in quantities not exceeding 
certain values were identified as being outside of the scope of the 
transportation regulations.
    The NRC has established an exemption at 10 CFR 71.14 that exempts 
licensees from the requirements of 10 CFR part 71 for certain natural 
materials and ores. The exemption for low-level materials exempts 
licensees from the requirements of 10 CFR part 71 with respect to the 
shipment or carriage of material that qualifies for the exemption and 
they would be allowed to transport natural material or ore that 
qualifies for the exemption without the material being regulated as a 
hazardous material during transportation; however, all other NRC 
regulations that apply to this material would continue to apply. The 
exemption at Sec.  71.14(a)(1) is consistent with the 1996 edition of 
TS-R-1 (as amended in 2000) and 49 CFR 173.401(b), as they apply to 
natural materials and ores containing naturally occurring 
radionuclides. The NRC is proposing to update this exemption to include 
the shipment of natural materials and ores containing naturally 
occurring radionuclides that have been processed, which would retain 
consistency with DOT regulations and harmonize the NRC regulations with 
the

[[Page 28993]]

2009 edition of TS-R-1. This exemption would continue to be limited to 
those natural materials and ores containing naturally occurring 
radionuclides whose activity concentrations may be up to 10 times the 
activity concentration specified in Table A-2 of appendix A to 10 CFR 
part 71.
    The NRC is proposing to correct the definition of LSA-I material, 
so that it applies to uranium and thorium ores, concentrates of uranium 
and thorium ores, and other ores containing naturally occurring 
radionuclides that are intended to be processed for their 
radionuclides. The low-level material exemption at Sec.  71.14(b)(3), 
which includes packages containing only LSA material, would now apply 
to LSA-I material (i.e., material intended to be processed for its 
radionuclides).
    Natural material and ore containing naturally occurring 
radionuclides that are not intended to be processed for these 
radionuclides could qualify for the low-level material exemption at 10 
CFR 71.14(a)(1). With the correction to the definition of LSA-I 
material, uranium and thorium ores, concentrates of uranium and thorium 
ores, and other ores containing naturally occurring radionuclides that 
are intended to be processed for these radionuclides may be able to 
qualify for the low-level material exemption at Sec.  71.14(b)(3), 
provided that the other restrictions are satisfied. The restrictions 
include: (1) the package contains only LSA-I or Surface Contaminated 
Object (SCO)-I material or (2) that the LSA or SCO material has an 
external radiation dose rate of less than 10 mSv/h (1 rem/h) at a 
distance of 3 meters from the unshielded material. Section 71.14 
provides an exemption from the requirements of 10 CFR part 71, with the 
exception of Sec. Sec.  71.5 and 71.88. Section 71.5 references the DOT 
regulations in 49 CFR parts 107, 171 through 180, and 390 through 397. 
If the DOT regulations are not applicable to a shipment of licensed 
material, Sec.  71.5 requires licensees to conform to the referenced 
DOT standards and regulations to the same extent as if the shipment 
were subject to the DOT regulations. Section 71.88 would continue to 
apply to the material, because its applicability is not limited by any 
of the exemptions in 10 CFR part 71.
    Natural material or ore that has been incorporated into a 
manufactured product, such as an article, instrument, component of a 
manufactured article or instrument, or consumer item, would not be able 
to qualify for the low level material exemption for natural materials 
and ores containing naturally occurring radionuclides. Slags, sludges, 
tailings, residues, bag house dust, oil scale, and washed sands that 
are the byproducts of processing or refining are examples of natural 
material or ore that has been processed and that may still qualify for 
the exemption, provided that the processed material has not been 
incorporated into a manufactured product.
    The NRC is proposing to add a definition of contamination and to 
expand the exemption at Sec.  71.14 to include non-radioactive solid 
objects with substances present on any surface not exceeding the levels 
used to define contamination. The derived values used in the definition 
of contamination are conservative with respect to transportation, and 
quantities of radioactive substances below these values would result in 
small amounts of exposure during normal conditions of transportation 
and would contribute to insignificant exposures under accident 
conditions. Contamination would be defined as quantities in excess of 
0.4 Bq/cm\2\ (1 x 10-5 [micro]Ci/cm\2\) for beta and gamma 
emitters and low toxicity alpha emitters, or 0.04 Bq/cm\2\ (1 x 
10-6 [micro]Ci/cm\2\) for all other alpha emitters.

E. How might the qualification of special form radioactive material 
change?

    The NRC is proposing to update the alternate tests in Sec.  71.75 
that may be used for the qualification of special form radioactive 
material to tests in more recent editions of the consensus standards. 
The NRC is proposing to incorporate by reference the Class 4 and Class 
5 impact tests and the Class 6 temperature test prescribed in the ISO 
document ISO 2919:1999(E). The NRC is proposing to incorporate by 
reference the leaktightness tests specified in ISO document 
9978:1992(E). The IAEA has adopted, in TS-R-1, the Class 4 and Class 5 
impact tests in ISO 2919:1999(E), the Class 6 temperature test in ISO 
2919:1999(E), and the leaktightness tests in ISO 9978:1992(E).
    The Class 4 impact test in ISO 2919:1999(E) would replace the 
impact test in Sec.  71.75(d)--the Class 4 impact test in ISO 2919, 
``Sealed Radioactive Sources--Classification,'' first edition (1980)--
and would be available for use with specimens that have a mass that is 
less than 200 grams. The Class 5 impact test, which is being added, 
would allow use of an ISO impact test for specimens that have a mass 
that is less than 500 grams. The updated ISO impact tests maintain the 
requirement that the mass of the hammer used in the test is greater 
than 10 times the mass of the specimen.
    The Class 6 temperature test in ISO 2919:1999(E) would replace the 
temperature test in Sec.  71.75(d)--the Class 6 temperature test in ISO 
2919, ``Sealed Radioactive Sources--Classification,'' first edition 
(1980). The Class 6 temperature test in ISO 2919:1999(E) is more 
stringent than the test that it replaces, because it requires the same 
specimen to be used for both portions of the temperature test. The 
Class 6 temperature test would continue to be more stringent than the 
testing required by Sec.  71.75(b).
    The leaktightness tests prescribed in ISO 9978:1992(E) would 
replace the tests in ISO/TR 4826, ``Sealed Radioactive Sources--Leak 
Test Methods,'' (1979). The consensus standard ISO 9978:1992(E) has 
replaced ISO/TR 4826:1979(E), which has been withdrawn by ISO. The NRC 
has determined that the leaktightness tests prescribed in ISO 
9978:1992(E) provide an equivalent level of radiological safety as the 
leaching assessment procedure in Sec.  71.75(c).
    The NRC is proposing to revise the definition of special form 
radioactive material to allow material tested using the current 
requirements to continue to be treated as special form material, 
provided that the testing was completed before the effective date of 
the final rule. This would allow material tested using requirements in 
effect at the time of the testing to continue to be used. The NRC is 
proposing to correct the reference to the version of Sec.  71.4 in the 
CFR that was in effect on March 31, 1996, by changing the date of the 
revision from January 1, 1983, to January 1, 1996.
    The NRC is proposing to replace ``edges'' with ``edge'' to describe 
the billet used for the percussion test in Sec.  71.75(b)(2). The edge 
corresponds to the circular edge at the face of the billet. This is 
intended to clarify the description of the billet and to maintain 
consistency with the language used by the DOT in 49 CFR 173.469.

F. What changes may be made to Appendix A, ``Determination of 
A1 and A2 Values,'' part 71 of Title 10 of the 
Code of Federal Regulations (CFR) ?

    The NRC is proposing the following changes to appendix A.
1. Determining the Quantity of Radioactive Material That Can Be Shipped 
in a Package That Contains Both Special Form and Normal Form 
Radioactive Material
    The NRC is proposing to specifically address how to calculate the 
limit of the activity that may be transported in a Type A package, if 
the package contains both special form and normal form

[[Page 28994]]

radioactive material and the identities and activity limits for the 
radionuclides are known. By including this equation, the NRC would 
increase the consistency between 10 CFR part 71 and TS-R-1 and would 
provide additional clarity on how to address cases where a package will 
contain both special form and normal form material. The equation is 
similar to those already used in 10 CFR part 71 for mixtures of special 
form material and mixtures of normal form material.
2. Table A-1, ``A1 and A2 Values for 
Radionuclides''
    The NRC is proposing to revise Table A-1 to make the values in 10 
CFR part 71 consistent with the values in Table 2, ``Basic radionuclide 
values,'' in TS-R-1. Specifically, the NRC is proposing to--add an 
entry for Kr-79, which has been added to Table 2 in the 2009 edition of 
TS-R-1; adopt the A1 and A2 values for Cf-252; 
revise footnote a to include the list of parent radionuclides whose 
A1 and A2 values include contributions from 
daughter radionuclides with half-lives of less than 10 days; and move 
and revise footnote c, which applies to Ir-192, so that the footnote 
applies only to Ir-192 in special form material.
    The A1 and A2 values are used for determining 
what type of package must be used for the transportation of radioactive 
material. The A1 values are the maximum amount of special 
form material allowed in a Type A package. The A2 values are 
the maximum activity of ``other than special form'' material allowed in 
a Type A package. A1 and A2 values are also used 
for several other packaging limits throughout TS-R-1, such as 
specifying Type B package activity leakage limits, low-specific 
activity limits, and excepted package contents limits. The values of 
A1 and A2 have been adopted in 10 CFR part 71 and 
are specified in appendix A.
    The IAEA has added an entry for Kr-79 in the Table 2 of the 2009 
edition of TS-R-1. The NRC is proposing to adopt these radionuclide-
specific values for Kr-79 in Table A-1. The radionuclide-specific 
values would replace the generic values in Table A-3, which are 
currently used for Kr-79. The radiological criteria underlying the 
A1 and A2 values for Kr-79 have not changed, but 
the radionuclide-specific values were derived using radionuclide-
specific information and better reflect the radiological hazard of Kr-
79 than the generic values that they would replace.
    The IAEA has revised the A1 value for Cf-252 to the 
value that currently applies to domestic transportation. In the 2004 
final rule for 10 CFR part 71 (69 FR 3698; January 26, 2004), the NRC 
did not adopt the A1 value for Cf-252 in TS-R-1 for domestic 
transportation, because the NRC was aware that the IAEA was considering 
changing the value back to the value that has been in 10 CFR part 71; 
the IAEA has subsequently made this change. The NRC is proposing to 
adopt the A1 value for Cf-252, which would apply to both 
international and domestic transportation, and to adopt the IAEA value 
for A2. The NRC is proposing to delete the A2 
value that applies only to domestic transportation. Making this change 
would improve the harmonization of 10 CFR part 71 with TS-R-1 by 
adopting the A2 value for Cf-252 in TS-R-1. Because the 
A2 value for Cf-252 was established by the IAEA using the Q-
system and current data for Cf-252, the A2 value for Cf-252 
would be consistent with the other values derived using the Q-system 
that has been incorporated into 10 CFR part 71.
    The NRC is proposing to revise footnote a to Table A-1 to identify 
the A1 and A2 values that include contributions 
from daughter radionuclides that have a half-life that is less than 10 
days. The proposed list corresponds to the radionuclides listed in 
footnote a to Table 2 in TS-R-1, with the exception of argon-42 (Ar-42) 
and tellurium-118 (Te-118). Ar-42 and Te-118 would not be included, 
because they do not appear within Table A-1.
    The NRC is proposing to revise footnote c to Table A-1 to make 
clear that the activity of Ir-192 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 is appropriate for Ir-192 in 
special form.
3. Table A-2, ``Exempt Material Activity Concentrations and Exempt 
Consignment Activity Limits for Radionuclides''
    The NRC is proposing to revise Table A-2 to make the values in 10 
CFR part 71 consistent with the values in TS-R-1 and to add an entry 
for Kr-79, which has been added to Table 2, ``Basic radionuclide 
values,'' in the 2009 edition of TS-R-1. The NRC is also proposing to 
update the list of parent radionuclides and their progeny in footnote b 
to Table A-2 by removing the chains for the parent radionuclides 
cerium-134 (Ce-134), radon-220 (Rn-220), thorium-226 (Th-226), and U-
240 and adding the chain for the parent radionuclide silver-108m (Ag-
108m) to make the footnote consistent with footnote (b) in Table 2 of 
TS-R-1. The NRC is proposing to update the activity limit for exempt 
consignment for Te-121m to match the values in TS-R-1.
    Material that has an activity concentration that is less than the 
activity concentration for exempt material would pose a very low 
radiological risk. The activity limit for exempt consignment has been 
established for the transportation of material in quantities small 
enough for which the total activity is unlikely to result in any 
significant radiological exposure. This would be the case even for 
material that exceeds the activity concentration for exempt material.
    Krypton-79 is not listed in Table A-2, and the values from Table A-
3, ``General Values for A1 and A2,'' in appendix 
A are used to determine the activity concentration for exempt material 
and the activity limit for exempt consignment for Kr-79. Radionuclide-
specific values for the activity concentration for exempt material and 
the activity limit for exempt consignment have been derived for Kr-79 
and are included in the 2009 edition of TS-R-1.
    In the 2005 edition of TS-R-1, the IAEA revised the activity limit 
for exempt consignment for Te-121m. The change to the activity level 
for exempt consignment for Te-121m, which is based on new analyses and 
information, is consistent with the objectives of the exemption values. 
Also, to conform to International Commission on Radiological Assistance 
(ICRP) and IAEA changes, the activity limit for exempt consignment for 
Te-121m in Table A-2 is being changed from 1 x 10\5\ Bq (2.7 x 
10-6 Ci) to 1 x 10\6\ Bq (2.7 x 10-5 Ci).
    The IAEA has revised the list of parent radionuclides and their 
progeny included in secular equilibrium in footnote (b) to Table 2, 
``Basic radionuclide values'' in TS-R-1. This revision arose from the 
adoption of the nuclide-specific basic radionuclide values from the 
Basic Safety Standards (IAEA Safety Series No. 115, ``International 
Basic Safety Standards for Protection against Ionizing Radiation and 
for the Safety of Radiation Sources'' (1996)) for use in 
transportation. The list of parent radionuclides and their progeny was 
modified by adding the decay chain for Ag-108m and removing the decay 
chain for Ce-134, Rn-220, Th-226, and U-240. The list of parent 
radionuclides and their progeny included in secular equilibrium 
presented in footnote b to Table A-2 would be revised to be consistent 
with the changes to the list in TS-R-1.

[[Page 28995]]

4. Table A-3, ``General Values for A1 and A2''
    In the 2005 Edition of TS-R-1, the IAEA revised Table 2, ``Basic 
radionuclide values for unknown radionuclides or mixtures'' (Table 3 in 
the 2009 edition of TS-R-1). The table divides unknown radionuclides 
and mixtures into three groups, with a row for each group. The first 
column of each row provides a descriptive phrase for contents that are 
suitable for that group. The current descriptive phrases are: (1) 
``only beta or gamma emitting radionuclides are known to be present,'' 
(2) ``only alpha emitting nuclides are known to be present,'' and 3) 
``no relevant data are available.'' The NRC is proposing to adopt the 
descriptive phrases as revised by the IAEA in TS-R-1 in Table A-3.
    The descriptive phrase for the first group, ``only beta or gamma 
emitting radionuclides are known to be present,'' is not being changed.
    The phrase for the second group, ``only alpha emitting nuclides are 
known to be present,'' is being changed to ``alpha emitting nuclides, 
but no neutron emitters, are known to be present.'' The phrase for the 
third group, ``no relevant data are available,'' is being changed to 
``neutron emitting nuclides are known to be present or no relevant data 
are available.'' Some users have assigned alpha-emitting radionuclides 
that also emit beta particles or gamma rays to the third group, when it 
was intended that they be assigned to the second group. The change in 
the descriptive phrase for the second group is intended to reduce the 
confusion caused by the current phrase, because all alpha emitting 
radionuclides also emit other particles and/or gamma rays. The change 
in the descriptive phrase for the third group is intended to clarify 
that neutron-emitting radionuclides, or alpha emitters that also emit 
neutrons, such as Cf-252, Cf-254 and curium-248 (Cm-248), should be 
assigned to the third group.
    It is intended that when groups of radionuclides are based on the 
total alpha activity and the total beta and gamma activity, the lowest 
radionuclide values (A1 or A2) for the alpha 
emitters or the beta or gamma emitters, respectively, would be used. 
Consequently, an A1 value of 1 TBq (2.7 Ci) and an 
A2 value of 9 x 10-5 TBq (2.4 x 10-3 
Ci) would be used for a group containing both alpha emitting 
radionuclides and beta or gamma emitting radionuclides.
5. Other changes that correct formulas and their descriptions in 
Section IV, Section-by-Section Analysis, of this document
    The NRC is proposing to make several corrections to the formulas 
and the descriptions of the formulas that address mixtures of 
radionuclides in Section IV of this document. These changes involve 
formatting and typographical changes in the formulas and their 
descriptions.

G. How would the responsibilities of certificate holders and licensees 
change with these amendments?

    In the 1950s, the Atomic Energy Commission (AEC) issued package 
approvals to AEC licensees as amendments to their licenses and the DOT 
issued package approvals to non-AEC licensees. On March 22, 1973 (38 FR 
8466), the AEC and the DOT entered into an MOU where the DOT agreed to 
adopt a requirement for AEC approval of designs of packages for the 
shipment of fissile material and other radioactive material exceeding 
Type A limits, with the exception of LSA material, and the AEC agreed 
to develop safety standards for the design and performance of packages 
and to impose these standards on AEC licensees and license-exempt 
contractors. Under the MOU, the AEC would issue an AEC license, an AEC 
CoC, or other AEC package approval directly to the person requesting 
the evaluation. Although the AEC, and subsequently the NRC, certified 
that the packages met the regulations, they did not have regulatory 
authority over the certificate holders under DOT jurisdiction. On July 
2, 1979 (44 FR 38690), this MOU was superseded by an MOU between the 
DOT and the NRC. In this MOU, it was agreed that the NRC, in 
consultation with the DOT, would develop safety standards for the 
design and performance of the packages. As the NRC developed its safety 
standards for the packages, it gained regulatory authority over the 
certificate holders.
    The requirements for making the preliminary determinations have 
remained largely unchanged since the 1979 MOU. In discussing the 
routine and preliminary determinations (48 FR 35600; August 5, 1983), 
the Commission indicated that the user of a package always had the 
regulatory responsibility for preliminary and routine determinations 
and recordkeeping, even though the user may not own the package. The 
Commission also indicated that although the user could contract with 
some other person, perhaps the owner, to satisfy those requirements for 
the user, the user's records must demonstrate that the requirements 
have been satisfied. Although leaktightness tests related to the 
package design are required as a condition of the package design 
approval, the Commission has indicated that it considers that in the 
case of radioactive material packages, integrity of the containment 
(including closures, valves, and other routes of escape) should be 
demonstrated for each fabricated package before first use.
    The NRC experience is that licensees have never made preliminary 
determinations themselves, unless they also happened to be certificate 
holders. Based on the NRC extensive experience inspecting the 
activities of certificate holders and NRC licensees who use packages, 
the NRC is not aware of any NRC licensee that performs preliminary 
determinations, unless they are also the certificate holder for the 
package design. The scope of user-only quality assurance program 
approvals, which are issued to licensees who are not also holders of a 
CoC, do not include the testing required to make the preliminary 
determinations. Licensees lease or buy these packages from the 
certificate holder, or fabricator, and most packages are already marked 
by the certificate holder. The NRC has identified cases where the 
durable marking of the packaging required by Sec.  71.85 was done 
incorrectly by a certificate holder. Because the licensee is 
responsible for the preliminary determinations, enforcement could not 
be taken against the certificate holder for improperly marking the 
packaging.
    The Commission is proposing to make changes to Sec.  71.85 that 
would make certificate holders, not licensees, responsible for making 
the preliminary determinations before the first use of each package. 
The preliminary determinations involve evaluating, testing, and marking 
the packaging. The DOT requirements at 49 CFR 173.22 require that the 
person offering a hazardous material for shipping make determinations 
relating to the manufacturing, assembly, and marking of the packaging 
or container. The Commission is proposing to require the licensee to 
ascertain that the preliminary determinations involving evaluating, 
testing, and marking the packaging have been made. The licensee would 
still make the required routine determinations at Sec.  71.87. As 
required by Sec.  71.91(d), both licensees and certificate holders 
would still be required to maintain sufficient written records to 
furnish evidence of the quality of the packaging, which includes the 
results of the determinations required by Sec.  71.85.

[[Page 28996]]

    The Commission is proposing to make these changes, because it is 
more appropriate to assign the responsibility to certificate holders 
for marking the packaging. Only certificate holders are authorized to 
design and fabricate packagings, and only certificate holders would 
have a full scope quality assurance program approval, which would allow 
them to perform the testing required as part of the preliminary 
determinations under an approved quality assurance program. However, 
licensees would need to retain their responsibility to determine that 
the packaging has been manufactured, assembled, and marked 
appropriately and that the packaging does not have any defects that 
could significantly reduce the effectiveness of the packaging. By 
assigning the responsibility for making the determinations to the 
certificate holder, the NRC would be able to streamline the 
implementation of its regulations and have the regulations better 
reflect current practice.

H. Why would renewal of my quality assurance program description not be 
necessary?

    The duration of quality assurance program approvals issued under 10 
CFR part 71 is a matter of practice and is not specified in the 
regulations. The NRC has limited the duration of the quality assurance 
program approval to provide an opportunity for the NRC staff to 
periodically review the quality assurance programs and for the NRC to 
maintain periodic contact with the quality assurance program approval 
holders. The limited duration of the approval facilitated the NRC 
recordkeeping relating to points of contact, package fabrication, use 
activities, and other administrative activities.
    In 2004, the NRC extended the duration of its quality assurance 
program approvals from 5 years to 10 years, because the NRC had 
determined that the periodic contact associated with the 5-year renewal 
period was less important than it was previously, and the duration of 
the approval could be lengthened. The NRC announced this change in RIS 
2004-18, ``Expiration Date for 10 CFR Part 71 Quality Assurance Program 
Approvals'' (December 1, 2004).
    The NRC is changing its practice regarding the duration of its 
quality assurance program approvals. The NRC would no longer limit the 
duration of its quality assurance program approvals issued under 10 CFR 
part 71. The NRC is proposing changes to 10 CFR part 71 to implement 
this change and to enhance the periodic communication between the NRC 
and the quality assurance program approval holders. The NRC would 
reissue its quality assurance program approval for Radioactive Material 
Packages (NRC Form 311) without an expiration date. As discussed in 
Section III, question I, ``What Changes Could be Made to a Quality 
Assurance Program Description without Seeking Prior NRC Approval?,'' 
and question J, ``How Frequently Would I Submit Periodic Updates on My 
Quality Assurance Program Description to the NRC?,'' the NRC is 
proposing to require quality assurance program approval holders to 
periodically report changes in their quality assurance program 
description to the NRC. The NRC has determined that with the continuing 
contact between the NRC and the quality assurance program approval 
holders, requiring the renewal of quality assurance program approvals 
is not necessary to provide the NRC with assurance that the quality 
assurance program approval holders would continue to be able to 
adequately maintain and implement their approved quality assurance 
program.
    As discussed under question I, ``What changes could be made to a 
quality assurance program description without seeking prior NRC 
approval?,'' the NRC would continue to approve quality assurance 
program description changes that reduce commitments made to the NRC in 
quality assurance program descriptions that have been approved by the 
NRC. Every 24 months, each quality assurance program approval holder 
would be required to report those changes that do not reduce 
commitments made to the NRC in a quality assurance program description 
approved by the NRC. Holders of a CoC and applicants for a CoC are 
subject to periodic inspection of their quality assurance program 
(approximately every 3 years) by the NRC. Licensees who use packages 
are inspected on an as-needed basis.
    As discussed under question P, ``What should I consider as I 
prepare my comments to the NRC?,'' the NRC is specifically requesting 
comment on the proposed approach to reporting changes to approved 
quality assurance program descriptions.

I. What changes could be made to a quality assurance program 
description without seeking prior NRC approval?

    Currently, quality assurance program descriptions approved under 10 
CFR part 71 cannot be changed without NRC approval. Therefore, all 
changes to 10 CFR part 71 quality assurance programs, irrespective of 
their significance or importance to safety, must be submitted to the 
NRC for approval. Licensees with quality assurance programs approved 
under 10 CFR part 50, may make some changes to their quality assurance 
program without NRC approval, consistent with the requirements at Sec.  
50.54. The NRC is proposing to allow some changes to be made to quality 
assurance programs approved under 10 CFR part 71 without obtaining NRC 
approval. The process for making changes to approved quality assurance 
program descriptions would be similar to the process that the NRC has 
used to approve changes that are made to the quality assurance program 
descriptions for nuclear power plants licensed under 10 CFR part 50 
through the provisions at Sec.  50.54(a) and would result in a more 
consistent approach to allowing changes to approved quality assurance 
programs. The NRC is proposing to establish a process that would 
require NRC approval to be obtained for those changes that are most 
important to safety but would allow other changes to be implemented 
without obtaining NRC approval.
    Quality assurance program approval holders would be required to 
obtain NRC approval before making any change to their quality assurance 
program description that would reduce the commitments that they have 
made to the NRC. Quality assurance program approval holders would not 
be required to submit changes to their quality assurance program 
descriptions, if those changes do not reduce the commitments that they 
have made to the NRC. Administrative changes (e.g., revisions to 
format, font size or style, paper size for drawings and graphics, or 
revised paper color) and clarifications, spelling corrections, and non-
substantive editorial or punctuation changes would not require NRC 
approval. Changes to reporting responsibilities, functional 
responsibilities, functional relationships, and some editorial or 
punctuation changes may be substantive and have the potential to reduce 
commitments made to the NRC and, in these instances, would require 
prior NRC approval before being implemented. The following includes 
types of changes that the NRC would not consider as reducing a 
commitment made to the NRC:
    1. The use of a quality assurance standard approved by the NRC, 
which is more recent than the quality assurance standard in the current 
quality assurance program at the time of the change;
    2. The use of generic organizational position titles that clearly 
denote the function of the position, supplemented

[[Page 28997]]

as necessary by descriptive text, rather than specific titles, provided 
that there are no substantive changes to either the functions of the 
position or reporting responsibilities;
    3. The use of generic organizational charts to indicate functional 
relationships, authorities, and responsibilities, or alternatively, the 
use of descriptive text;
    4. The elimination of quality assurance program information that 
duplicates language in quality assurance regulatory guides and quality 
assurance standards to which the holder of the quality assurance 
program approval has committed on record; and
    5. Organizational revisions that ensure that persons and 
organizations performing quality assurance functions continue to have 
the requisite authority and organizational freedom, including 
sufficient independence from cost and schedule when opposed to safety 
considerations.
    Quality assurance program approval holders would also need to 
maintain records of all quality assurance program changes.

J. How frequently would I submit periodic updates on my quality 
assurance program description to the NRC?

    The NRC would continue to require quality assurance program 
approval holders to obtain NRC approval of any change to their approved 
quality assurance program description that would reduce any commitment 
in the quality assurance program description approved by the NRC before 
they implement the change. The NRC would require the following 
information to be provided for its review: a description of the 
proposed changes to the approved quality assurance program description, 
the reason for the change, and the basis for concluding that the 
revised program incorporating the change continues to satisfy the 
requirements of subpart H.
    The NRC is proposing to require that quality assurance program 
approval holders would report changes to their approved quality 
assurance program that do not reduce any commitments in the quality 
assurance program description approved by the NRC every 24 months. 
These changes would not require NRC approval before they can be 
implemented. If the quality assurance program approval holder has not 
made any changes to its approved quality assurance program description 
during the preceding 24-month period, it would report to the NRC that 
no changes have been made.
    The NRC inspection program relies on having current information 
about the quality assurance program available to the NRC. By requiring 
that the most important changes be submitted to the NRC before they are 
implemented and with the periodic reporting of the less significant 
changes every 24 months, the NRC would have current information for its 
inspection program. The NRC considers the 24-month reporting period as 
providing an appropriate balance between the burden placed on the 
quality assurance program approval holders and the need to ensure that 
the NRC has current information for its oversight of these quality 
assurance programs.
    As discussed under question H, ``Why would renewal of my quality 
assurance program description not necessary?,'' the NRC would re-issue 
NRC Form 311 without an expiration date. The 24-month period for 
reporting of changes is proposed to begin on the date of the NRC 
approval of a quality assurance program issued with no expiration date, 
as specified by the date of signature at the bottom of NRC Form 311, 
``Quality Assurance Program Approval for Radioactive Material 
Packages.''
    As discussed under question P, ``What should I consider as I 
prepare my comments to the NRC?,'' the NRC is proposing to require 
quality assurance program approval holders to submit a report every 2 
years that describes the changes that were made to their quality 
assurance program description that do not reduce a commitment in the 
quality assurance program description approved by the NRC. The NRC is 
seeking to balance the regulatory burden for submitting this 
information with the NRC need to ensure that the NRC has current 
information for its regulatory oversight of quality assurance program 
approval holders, which would include using the information for 
inspections. The NRC is requesting comment on the following issue: 
would a different frequency be more appropriate for reporting changes 
to approved quality assurance programs that do not reduce a commitment 
in a quality assurance program description approved by the NRC?

K. How would the requirements in subpart H, ``Quality Assurance,'' 
change with the removal of the footnote in 10 CFR 71.103?

    The NRC is proposing to remove the footnote in Sec.  71.103 
regarding the use of the term ``licensee'' in subpart H, because it is 
no longer necessary. The removal of the footnote does not change the 
quality assurance requirements in subpart H. The footnote regarding use 
of the term ``licensee'' was included to clarify that the quality 
assurance requirements in subpart H apply to whatever design, 
fabrication, assembly, and testing of a package is accomplished before 
a package approval is issued. The terms ``certificate holder'' and 
``applicant for a CoC'' were added to the requirements in subpart H in 
a later rulemaking to make explicit the application of those quality 
assurance requirements to certificate holders and applicants for a CoC. 
Although removing the footnote would not change the quality assurance 
requirements, other proposed changes to subpart H in this proposed 
rulemaking would further clarify which requirements apply to users of 
NRC certified packaging and which apply to applicants for, or holders 
of, CoCs--the entities that would be performing design, fabrication, 
assembly, and testing of the package before a package approval is 
issued.

L. What changes would be made to general licenses?

    The NRC is proposing to change the requirements for general 
licenses for the following: (1) use of an NRC-approved package (Sec.  
71.17) and 2) use of a foreign-approved package (Sec.  71.21). In Sec.  
71.17, the NRC is revising the general license requirements to clarify 
the conditions for obtaining a general license and the responsibilities 
of the general licensee. A quality assurance program approved by the 
Commission as satisfying the provisions of subpart H of 10 CFR part 71 
is required to be granted the general license. The proposed changes 
would clarify that the licensee is responsible for maintaining copies 
of the appropriate documents, such as the CoC, or other approval of the 
package, and the documents associated with the use and maintenance of 
the packaging and the actions that are to be taken before shipment with 
the package. The changes would also clarify that making the 
notification in Sec.  71.17(c)(3) to the NRC is a responsibility of the 
licensee, rather than a condition for obtaining the license. The 
proposed changes to Sec. Sec.  71.17 and 71.21 would not change the 
current notification process and would not change the required timing 
or content of the notification required by Sec.  71.17(c)(3) or any 
other reporting requirements relating to package use or, where 
required, the prior notification of shipments.
    The proposed changes also include updating the reference in Sec.  
71.21(a) from 49 CFR 171.12 to 49 CFR 171.23. On May 3, 2007 (72 FR 
25162), the DOT published a final rule that moved the requirements at 
49 CFR 171.12 to paragraph (b)(11) at 49 CFR 171.23, ``Requirements for 
the specific materials

[[Page 28998]]

and packagings transported under the [International Civil Aviation 
Organization] ICAO Technical Instructions, [International Maritime 
Dangerous Goods] IMDG Code, Transportation Canada [Transportation of 
Dangerous Goods] TDG Regulations, or the IAEA Regulations.''

M. How would the exemption from classification as fissile material (10 
CFR 71.15) change?

    The objective of the fissile material exemptions at Sec.  71.15 is 
to facilitate the safe transport of low-risk (e.g., small quantities or 
low concentrations) of fissile material by exempting shipments of these 
materials from the packaging requirements and the criticality safety 
assessments required for fissile material transportation and to allow 
the shipments to take place without specific Commission approval. The 
lower amount of regulatory oversight is acceptable for these shipments, 
because the exemptions are established so as to ensure safety under all 
credible transportation conditions. Provided that the exempt material 
is packaged consistent with the radioactive and hazardous properties of 
the material, there would not be any additional packaging or transport 
requirements for exempt fissile material beyond that noted in the 
specific exemption. However, exempt fissile material would still have 
fewer restrictions imposed than if it were to be shipped as fissile 
material. Therefore, for purposes of ensuring criticality safety, the 
exemptions consider that the material can be released from any 
packaging during transport, may reconfigure into a worst-case geometric 
arrangement, may combine with material from other transport vehicles, 
and may be subject to the fire and water immersion conditions assumed 
as part of the criticality safety assessment for package designs 
approved to transport fissile material.
    The reactivity of uranium enriched in U-235 will depend on the 
level of enrichment, the presence of moderators, and heterogeneity 
effects. Hydrogen is the most efficient moderator, and water is the 
most common material containing large quantities of hydrogen; 
therefore, water is the typical moderating material of interest in 
criticality safety. The maximum enrichment in U-235 allowed to qualify 
for the fissile material exemption at Sec.  71.15(d) is 1 percent by 
weight, which is slightly less than the minimum critical enrichment for 
an infinite, homogeneous mixture of enriched uranium and water.\3\ The 
minimum critical enrichment is the enrichment necessary for a system to 
have a neutron multiplication factor of one. Systems containing 
homogeneous mixtures of uranium enriched to less than the minimum 
critical enrichment (e.g., a homogenous mixture of uranium enriched to 
a maximum one percent) will not be critical, irrespective of the mass 
or size of the system. The fissile material exemption at Sec.  71.15(d) 
also limits the quantity of some less common moderating materials 
(beryllium, graphite, hydrogenous material enriched in deuterium), 
because the presence of these materials has the potential to reduce the 
minimum critical enrichment, increasing the potential for criticality 
with uranium of lower enrichment. Thus, homogeneous materials 
containing uranium enriched to no more than 1 percent by weight and 
subject to the noted restrictions on moderators will be inherently safe 
from a potential criticality, because they do not need to be limited by 
mass or size to be subcritical during transport. However, uranium 
enriched to less than 5 percent by weight is most reactive when it is 
in a heterogeneous configuration; therefore, the minimum critical 
enrichment would be lower for an optimized heterogeneous system than 
for an optimized homogeneous system of the same material. In 
consideration of this fact, the current proposed change at Sec.  
71.15(d) is to add requirements to clarify the need for homogeneity in 
the material.
---------------------------------------------------------------------------

    \3\ H.C. Paxton and N. L. Pruvost, Critical Dimensions of 
Systems Containing U-235, Pu-239, and U-233, LA-10860-MS, Los Alamos 
National Laboratory, (1987).
---------------------------------------------------------------------------

    The exemption for uranium enriched to a maximum of 1 percent at 
Sec.  71.15(d) includes a limit on moderators that increase the 
reactivity of the low-enriched fissile material, but the exemption does 
not include limits on heterogeneity. In contrast, TS-R-1 allows the 
uranium enriched to a maximum of 1 percent by weight to be distributed 
essentially homogeneously throughout the material and requires that if 
the U-235 is in metallic, oxide, or carbide forms, then it cannot form 
a lattice arrangement; however, TS-R-1 does not limit the amount of 
beryllium, graphite, or hydrogenous material enriched in deuterium. In 
its supplemental guidance to TS-R-1, ``Advisory Material for the IAEA 
Regulations for the Safe Transport of Radioactive Material'' (TS-G-
1.1), the IAEA indicated that ``[t]here is agreement that homogeneous 
mixtures and slurries are those in which the particles in the mixture 
are uniformly distributed and have a diameter no larger than 127 [mu]m 
[(5 x 10-3 in.)].'' The homogeneity requirement in TS-R-1 is 
intended to prevent latticing of slightly enriched uranium in a 
moderating medium.
    As described in Section II, Background, of this document, analyses 
performed by the DOE indicated that large arrays of uranium with 
enrichment of 1 percent by weight of U-235, which would qualify for the 
fissile material exemption at Sec.  71.15(d), could exceed an effective 
neutron multiplication factor (keff) of 0.95 when optimally 
moderated by water. The DOE analyses were performed assuming five 
shipments under normal conditions and two shipments under accident 
conditions. Shipping the material under the exemption would have 
resulted in a lower margin of safety with respect to criticality than 
is allowed for shipments using approved fissile material packages, 
because shipments using the fissile material packages, by design, would 
typically use a keff of 0.95 as an upper limit. Because such 
a shipment, as was analyzed by the DOE, could both qualify for the 
fissile material exemption for low-enriched fissile material and have a 
keff greater than 0.95, the Commission believes that 
additional restrictions on low-enriched fissile material shipped under 
the fissile material exemption at Sec.  71.15(d) are warranted.
    When the Commission last identified a defect in its fissile 
exemption regulations, which allowed shipments to be made without prior 
Commission approval, the Commission published an emergency final rule 
to restrict the use of beryllium and other special moderators, such as 
graphite and hydrogenous material enriched in deuterium. In this 
instance, the Commission chose to use normal notice-and-comment 
rulemaking procedures and determined that the proposed change did not 
need to be effective immediately. Uranium enriched to a maximum of 1 
percent by weight is rarely available in quantities that would allow 
keff to exceed 0.95. In the case of uranium enriched to a 
maximum of 1 percent by weight, keff is not sensitive to 
changes in mass, so a significant amount of additional mass would be 
required to increase the keff from 0.95 to a value very 
close to 1.0, even when geometry and moderator conditions are optimal 
with respect to criticality. In addition, keff is very 
sensitive to moderator conditions. If the moderator conditions are not 
optimal, keff is less sensitive to changes in mass. 
Therefore, it is very unlikely that even in the case of large

[[Page 28999]]

quantities of uranium enriched to a maximum of 1 percent by weight that 
the moderator conditions would also be close to optimal with respect to 
criticality. The upper subcritical limit is the maximum allowed value 
of keff and includes a minimum margin of subcriticality. At 
a keff equal to 1, the system is considered critical.
    As discussed in Section II of this document, the NRC removed both 
the requirement for uranium enriched to a maximum of 1 percent to be 
homogeneously distributed and the lattice prevention requirement. 
Although the NRC had determined that the limits on restricted 
moderators was sufficient to assure subcriticality for all moderators 
of concern, the NRC believes that additional restrictions are needed to 
have a sufficient margin of safety for shipments of material under the 
low-enriched fissile material exemption. Therefore, the NRC is 
proposing to reinstate the requirement that, for uranium enriched to a 
maximum of 1 percent to be exempted, the fissile material must be 
distributed homogeneously throughout the package contents and not form 
a lattice arrangement. Some variability in the distribution and 
enrichment of the uranium enriched to a maximum of 1 percent would be 
permissible, provided that the maximum enrichment does not exceed 1 
percent. The total measured mass of U-233 and plutonium, plus two times 
the measurement uncertainty, should be less than 1.0 percent of the 
mass of U-235 in the material. The total measured mass of beryllium, 
graphite, and hydrogenous material enriched in deuterium, plus two 
times the measurement uncertainty, should be less than 5.0 percent of 
the uranium mass. Although there are heterogeneity effects at very 
small scales, the Commission does not believe that it is necessary to 
require homogeneity with respect to particle size. Further, the 
Commission does not consider it to be credible to accumulate the volume 
and regularity of fissile material particles necessary for small-scale 
heterogeneity to introduce criticality concerns. Small volumes of 
heterogeneity may exist for material shipped under this exemption, 
provided that a significant fraction of the fissile material is 
homogeneous and mixed with non-fissile material, or the lumps of 
fissile material are spaced in a largely irregular arrangement. The 
homogeneity criterion--allowing some variability in the distribution of 
fissile material--is consistent with the IAEA regulations, which 
require that the fissile nuclides be essentially homogenously 
distributed. Restricting the variability in concentration is not 
sufficient for limiting the reactivity of the uranium enriched to a 
maximum of 1 percent. Therefore, the Commission is also proposing to 
reinstate the lattice prevention criterion. The contents of the package 
should not involve concentrations of fissile material separated by non-
fissile material in a regular, lattice-like arrangement. Although the 
lattice prevention requirement in TS-R-1 is limited to uranium present 
in metallic, oxide, or carbide form, the Commission believes that this 
restriction is too narrow and should apply irrespective of the form of 
uranium. As discussed under question P, ``What should I consider as I 
prepare my comments to the NRC?,'' the NRC is seeking comment on the 
homogeneity and lattice prevention requirements for the exemption for 
uranium enriched to a maximum of 1 percent. The Commission is 
requesting comment on the clarity of the homogeneity and lattice 
prevention criteria for implementation.

N. What other changes is the NRC proposing to make to its regulations 
for the packaging and transportation of radioactive material?

    A requirement in Sec.  71.19(a) that implemented transitional 
arrangements (``grandfathering'') expired on October 1, 2008, and has 
been deleted. Paragraph 71.19(a) is currently reserved. Other 
paragraphs in Sec.  71.19 would be redesignated. In redesignated 
paragraph 71.19(b)(2), transitional language that is no longer needed 
would be removed, because the transitional period has expired and the 
requirement now applies to all previously approved packages used for a 
shipment to a location outside of the United States.
    References to Sec.  71.20 in Sec.  71.0 would be removed, because 
Sec.  71.20 has expired and has been removed from the regulations.
    In Sec.  71.31, the reference to Sec.  71.13 would be changed to 
Sec.  71.19. In Sec.  71.91, the reference to Sec.  71.10 would be 
changed to Sec.  71.14. These changes would correct references that 
were not updated when the requirements were redesignated in 2004.
    In Sec.  71.101, the NRC is proposing to make changes that would 
make the requirements more precise. Paragraphs 71.101(a) and 
71.101(c)(2) would be revised to clarify the responsibilities of 
licensees and certificate holders and applicants for a CoC. The quality 
assurance requirements pertaining to the design, fabrication, testing, 
and modification of packaging apply to certificate holders and 
applicants for a CoC. Licensees are responsible for the quality 
assurance requirements that apply to their use of the packaging for the 
shipment of licensed material. Paragraph 71.101(c) would be changed to 
remove the overlap between paragraphs (c)(1) and (c)(2), by removing 
the reference to licensees in paragraph (c)(2).

O. When would these proposed amendments become effective?

    The NRC will coordinate the effective date for this rule with the 
DOT. As described under question P, ``What Should I Consider as I 
Prepare My Comments to the NRC?,'' the NRC is requesting comments on 
the cumulative effects of regulation (CER), including comments that 
would inform the amount of time that would be sufficient to implement 
the proposed amendments. The NRC intends that the new regulations would 
become effective no sooner than 90 days after the final rule is 
published in the Federal Register.

P. What should I consider as I prepare my comments to the NRC?

    Tips for preparing your comments--when submitting your comments, 
remember to:
    1. Identify the rulemaking (RIN 3150-AI11; NRC-2008-0198).
    2. Explain why you agree or disagree; suggest alternatives and 
substitute language for your requested changes.
    3. Describe any assumptions and provide any technical information 
and/or data that you used.
    4. If you estimate potential costs or burdens, explain how you 
arrived at your estimate in sufficient detail to allow for it to be 
reproduced.
    5. Provide specific examples to illustrate your concerns, and 
suggest alternatives.
    6. Explain your views as clearly as possible.
    7. Make sure to submit your comments by the comment period deadline 
identified.
    8. See Section VIII for the request for comments on the use of 
plain writing, Section IX for the request for comments on the adoption 
of voluntary consensus standards, Section XI for the request on the 
reporting and recordkeeping burden, and Section XII for the request for 
comments on the draft regulatory analysis.
    9. The NRC is specifically requesting comments on the following 
items:
    a. As discussed under question J, ``How frequently would I submit 
periodic updates on my quality assurance program to the NRC,'' the NRC 
is proposing to require quality assurance program approval holders to

[[Page 29000]]

submit a report every 2 years that describes the changes that were made 
to their quality assurance program that do not reduce a commitment in 
the quality assurance program description approved by the NRC. The NRC 
is seeking to balance the regulatory burden for submitting this 
information with the NRC need to ensure that the NRC has current 
information for its regulatory oversight of quality assurance program 
approval holders, which includes using the information for inspections. 
Inspections of certificate holders occur approximately every 3 years 
and inspections of licensees who use packages occur on an as-needed 
basis. The NRC is requesting comment on whether a different frequency 
would be more appropriate for reporting changes to an approved quality 
assurance program that do not reduce a commitment in a quality 
assurance program description approved by the NRC.
    b. In Sec.  71.15(d), the NRC is proposing to reintroduce 
restrictions on low-enriched fissile material--uranium enriched in U-
235 to a maximum of 1 percent by weight, and with a total plutonium and 
U-233 content of up to 1 percent of the mass of uranium-235--by 
requiring that it be distributed homogeneously and not form a lattice 
arrangement. The NRC is seeking comment on the clarity of this 
requirement for implementation.
    c. The CER describe the challenges that licensees, certificate 
holders, States, or other entities may encounter when implementing the 
new regulatory requirements (e.g., rules, generic letters, orders, 
backfits, inspections). The CER is an organizational effectiveness 
challenge that results from a licensee or impacted entity implementing 
a significant number of new or complex regulatory actions, within a 
limited implementation period and with available resources (which may 
include limited available expertise to address a specific issue). The 
CER can potentially distract licensee or other entity staff from 
executing other primary duties that ensure safety or security. The NRC 
is specifically requesting comment on the cumulative effects of this 
proposed rulemaking. In developing comments on the CER, consider the 
following questions:
    i. In light of any current or projected CER challenges, would the 
proposed rule's effective date provide sufficient time to implement the 
new proposed requirements, including changes to programs and 
procedures?
    ii. If current or projected CER challenges exist, what should be 
done to address this situation (e.g., if more time is required to 
implement the new requirements, what period of time would be 
sufficient)?
    iii. Do other (NRC or other agency) regulatory actions (e.g., 
orders, generic communications, license amendments requests, inspection 
findings of a generic nature) influence the implementation of the 
proposed requirements?
    iv. Are there unintended consequences? Does the proposed rule 
create conditions that would be contrary to the proposed rule's purpose 
and objectives? If so, what are the unintended consequences and how 
should they be addressed?
    v. Please comment on the NRC cost and benefit estimates in the 
regulatory analysis that supports the proposed rule.

IV. Section-by-Section Analysis

Section 71.0 Purpose and Scope

    Paragraph (d)(1) would be revised to delete Sec.  71.20 from the 
list of sections that a general license is issued without requiring the 
NRC to issue a package approval, so the reference to ``Sec. Sec.  71.20 
through 71.23'' would be revised to ``Sec. Sec.  71.21 through 71.23.''

Section 71.4 Definitions

    The definition of ``contamination'' would be added and would be 
consistent with the definition of contamination in DOT regulations at 
49 CFR 173 and TS-R-1.
    The definition of ``Criticality Safety Index (CSI)'' would be 
revised to be more consistent with the definition in DOT regulations at 
49 CFR 173 and TS-R-1 by addressing overpacks and freight containers in 
the definition.
    The definition of ``Low Specific Activity (LSA) material'' would be 
revised to be more consistent with the definition in DOT regulations at 
49 CFR 173 and TS-R-1 by revising paragraphs (1)(i) and (1)(ii). In 
paragraph (1)(i), the definition is changed to make the description of 
LSA-I material apply to material that is intended to be processed for 
the use of the uranium, thorium, and other naturally occurring 
radionuclides.
    The definition of ``Special form radioactive material'' would be 
revised to allow special form radioactive material that was 
successfully tested using the current requirements of Sec.  71.75(d) to 
continue to qualify as special form material, if the testing was 
completed before the date of the final rule. The reference to the 
version of 10 CFR part 71 in effect on March 31, 1996, would be 
corrected by changing 1983 to 1996.
    The definition of ``Uranium--natural, depleted, enriched'' would be 
revised by adding ``(which may be chemically separated)'' to paragraph 
(1), which applies to natural uranium.

Section 71.6 Information Collection Requirements: OMB Approval

    Paragraph (b) would be revised to add Sec.  71.106 to the list of 
sections with information collections.

Section 71.14 Exemption for Low-Level Materials.

    Paragraph 71.14(a)(1) would be revised to allow natural material 
and ores that contain naturally occurring radionuclides and that have 
been processed for purposes other than the extraction of the 
radionuclides to qualify for the exemption. Natural material or ore 
that has been processed, but has not been incorporated into a 
manufactured product, such as an article, instrument, component of a 
manufactured article or instrument, or consumer item could qualify for 
the exemption. Slags, sludges, tailings, residues, bag house dust, oil 
scale, and washed sands that are the byproducts of processing or 
refining would be considered as a natural material and could qualify 
for the exemption, provided that they were not incorporated into a 
manufactured product. To qualify for this exemption, the activity 
concentration of the natural material or ore could not exceed 10 times 
the activity concentration values and the material is not intended to 
be processed for the use of the radionuclides.
    A reference to Table A-3 in appendix A would be added in paragraphs 
71.14(a)(1) and (a)(2) as a source of activity concentration values 
that may be used to determine whether natural material or ore would 
qualify for the exemption. Table A-3 would provide activity 
concentration values for exempt material that would be used for 
individual radionuclides whose identities are known, but which are not 
listed in Table A-2.
    Paragraph 71.14(a)(3) would be added to provide an exemption for 
non-radioactive solid objects that have radioactive substances present 
on the surfaces of the object, provided that the quantity of 
radioactive substances is below the quantity used to define 
contamination. The definition of ``contamination'' would be added to 
Sec.  71.4.

Section 71.15 Exemption From Classification as Fissile Material

    Paragraph 71.15(d), which applies to fissile material in the form 
of uranium enriched in U-235 to a maximum of 1 percent by weight, would 
be revised.

[[Page 29001]]

The fissile material would be required to be distributed homogeneously 
and not form a lattice arrangement, where concentrated fissile material 
is separated by non-fissile material in a regular, repeating pattern.

Section 71.17 General License: NRC-Approved Package

    Paragraph 71.17(c) would be revised to clarify that the general 
licensee must comply with the requirements in Sec.  71.17(c)(1) through 
(c)(3).

Section 71.19 Previously Approved Package

    Paragraphs 71.19(b) through (e) would be redesignated as Sec. Sec.  
71.19(a) through (d).
    In redesignated Sec.  71.19(b)(2), the phrase ``[a]fter December 
31, 2003'' would be deleted. This would not change the requirement that 
packages used for a shipment to a location outside the United States 
would continue to be subject to multilateral approval as defined in the 
DOT regulations at 49 CFR 173.403, because all such shipments would 
occur after December 31, 2003.

Section 71.21 General License: Use of Foreign Approved Package

    Paragraph 71.21(a) would be revised to update the reference to 49 
CFR 171.12 to 49 CFR 171.23.
    Paragraph 71.21(d) would be revised to clarify that the general 
licensee must comply with the requirements in Sec.  71.21(d)(1) and 
(d)(2). Paragraph 71.21(d)(2) would be revised to delete the sentence 
regarding exemption from quality assurance provisions in subpart H for 
design, construction, and fabrication activities, because these 
requirements are not applicable to a general licensee. The general 
licensee would be required to comply with the quality assurance 
requirements in subpart H that do apply.

Section 71.31 Contents of Application

    In paragraph 71.31(b), the reference to ``Sec.  71.13'' would be 
corrected to ``Sec.  71.19.'' This change was inadvertently omitted 
during a previous rulemaking, when certain sections were renumbered.

Section 71.38 Renewal of a Certificate of Compliance

    The title of this section would be revised to remove the reference 
to the renewal of quality assurance program approvals. The section 
would be revised to be limited to the renewal of CoCs by removing all 
references to quality assurance program approvals. The NRC is changing 
its practice regarding the duration of quality assurance program 
approvals. Quality assurance program approvals would not have an 
expiration date, and the NRC would revise the current quality assurance 
program approvals so that they would not have an expiration date. The 
renewal of a quality assurance program approval would be unnecessary. 
Paragraph 71.38(c) would also be revised for improved clarity.

Section 71.70 Incorporations by Reference

    This section would be added to incorporate by reference the 
consensus standards referenced in Sec.  71.75--ISO 9978:1992(E), 
``Radiation protection--Sealed radioactive sources--Leakage test 
methods'' and ISO 2919:1999(E), ``Radiation protection--Sealed 
radioactive sources--General requirements and classification''--and 
would describe the availability of the documents.

Section 71.75 Qualification of Special Form Radioactive Material

    In Sec.  71.75(a)(5), the 1992 edition of ISO 9978 would be 
incorporated by reference for the alternate leak test methods for the 
qualification of special form material. The ISO/TR 4826 has been 
withdrawn by ISO and replaced by ISO 9978. This change would make 10 
CFR part 71 consistent with the DOT requirements in 49 CFR 173, which 
incorporated ISO 9978:1992(E) in 2004.
    In Sec.  71.75(b)(2)(ii), the description of the billet used in the 
percussion test would be changed to provide better clarity and to 
maintain consistency with the language used by the DOT in 49 CFR 
173.469 by replacing ``edges'' with ``edge.'' The edge corresponds to 
the circular edge at the face of the billet.
    In Sec.  71.75(b)(2)(iii), the description of the sheet of lead 
used in the percussion test would be changed to correct the thickness 
of the sheet of lead used in the percussion test to indicate that the 
thickness must not be more than 25 mm (1 inch) thick to be consistent 
with the thickness in TS-R-1.
    In Sec.  71.75(d), Sec. Sec.  71.75(d)(1)(i) and (d)(1)(ii) would 
be added. In Sec.  71.75(d), the 1999 edition of ISO 2919 would be 
incorporated by reference, replacing the reference to the 1980 edition 
of ISO 2919 for the alternate Class 4 impact test in Sec.  
71.75(d)(1)(i) and the alternate Class 6 temperature test in Sec.  
71.75(d)(2). The availability and other language incorporating this 
standard by reference is moved to Sec.  71.70. Paragraph 
71.75(d)(1)(ii) would allow the Class 5 impact tests prescribed in the 
1999 edition of ISO 2919 to be used in place of the impact and 
percussion tests in Sec. Sec.  71.75(b)(1) and (b)(2), if the specimen 
weighs less than 500 grams.

Section 71.85 Preliminary Determinations

    In Sec.  71.75(a), (b), and (c), ``licensee'' would be replaced by 
``certificate holder.'' The NRC experience is that these determinations 
are performed by the certificate holders who manufacture the package. 
This change would make the requirements consistent with current 
practice, because only certificate holders would have a quality 
assurance program approval that would allow them to conduct the 
required tests under an approved quality assurance program. Paragraph 
71.85(d) would be added to address the responsibilities of licensees 
using a package for transportation. Although certificate holders would 
be required to make the preliminary determinations under Sec.  
71.85(a), (b), and (c), the licensee would be responsible for ensuring 
that these determinations have been made before their first use of the 
packaging.

Section 71.91 Records

    In Sec.  71.91(a), the reference to ``Sec.  71.10'' would be 
corrected to ``Sec.  71.14.'' This reference was not updated when Sec.  
71.10 was redesignated as Sec.  71.14.

Section 71.101 Quality Assurance Requirements

    Paragraph 71.101(a) would be changed to clarify that certificate 
holders and applicants for a package approval are responsible for 
satisfying the quality assurance requirements that apply to design, 
fabrication, testing, and modification of packaging. The last two 
sentences would be revised to be more precise and to provide clarity.
    Paragraph 71.101(c)(2) would be changed to remove the reference to 
licensees in the first sentence. This would remove the overlap between 
the two paragraphs, by making it clear that licensees would notify the 
NRC before their first use of any package as required under Sec.  
71.75(c)(1) and certificate holders and applicants for a CoC would 
notify the NRC before the fabrication, testing, or modification of a 
package as required under Sec.  71.75(c)(2).

Section 71.103 Quality Assurance Organization

    In Sec.  71.75(a), footnote 2 would be removed. The activities 
described in the footnote are performed by certificate holders and 
applicants for a CoC. The footnote is unnecessary, because the 
requirements no longer rely on the use of the term ``licensee'' for 
those

[[Page 29002]]

activities performed by certificate holders and applicants for a CoC.

Section 71.106 Changes to a Quality Assurance Program

    This section would be added to establish requirements that would 
apply to changes to quality assurance programs. It would allow some 
changes to a quality assurance program to be made without obtaining the 
prior approval of the NRC. Currently, all changes, no matter how 
insignificant, must be approved by the NRC before they can be 
implemented. These provisions would allow changes to quality assurance 
programs that do not reduce commitments, such as those that involve 
administrative improvements and clarifications and editorial changes, 
to be made and implemented without NRC approval. Quality assurance 
program approval holders would be required to get NRC approval before 
making changes to their quality assurance program that would reduce 
their commitments to the NRC.
    Paragraph 71.106(a) would establish the requirements that would 
apply when a holder of a quality assurance program approval intends to 
make a change in its quality assurance program that would reduce their 
commitments to the NRC. The holder of a quality assurance program 
approval would be required to identify the change, the reason for the 
change, and the basis for concluding that the revised program 
incorporating the change would continue to satisfy the requirements of 
subpart H that apply.
    Paragraph 71.106(a)(2) would require that each holder of a quality 
assurance program approval maintain quality assurance program changes 
as records. These records would need to be maintained as required in 
Sec.  71.135.
    Paragraph 71.106(b) would allow the holder of a quality assurance 
program approval to make changes to its quality assurance program that 
would not reduce its commitments to the NRC and identifies the changes 
that would not be considered as reducing its commitments to the NRC.
    Paragraph 71.106(c) would require that records are maintained for 
any changes to the quality assurance program.

Section 71.135 Quality Assurance Records

    This section would be revised to include those quality assurance 
records that apply to changes that are made to approved quality 
assurance programs. The second sentence is revised to include the 
changes to the quality assurance program as required by Sec.  71.106 in 
the list of the types of records to be maintained.

Appendix A--Determination of A1 and A2.

    In paragraphs IV.a. through IV.f., the equations and accompanying 
text would be revised to make minor corrections. In paragraphs IV.a. 
and IV.b., the description of the equations would make it explicit that 
B(i) is the activity of radionuclide i in special form and normal form 
in paragraphs IV.a. and IV.b., respectively.
    Paragraph IV.c. would be added and paragraphs IV.c. through IV.f. 
would be redesignated as paragraphs IV.d. through IV.g., respectively. 
Paragraph IV.c. would provide an equation to be used for determining 
the quantity of radioactive material that can be shipped in a package 
that contains both special form and normal form radioactive material. 
This equation would increase the consistency between appendix A and TS-
R-1.
    In paragraph V., the existing text would be redesignated as 
paragraph V.a. Paragraph V.b. would be added to provide direction on 
calculating the exempt activity concentration for a mixture and the 
exempt consignment activity limit of a mixture, when the identity of 
each radionuclide is known, but the individual activities of some 
radionuclides are not known.
    Table A-1 would be revised to change the A1 value for 
Cf-252 from 5.0 x 10-2 TBq to 1.0 x 10-1 TBq, and 
from 1.4 Ci to 2.7 Ci. Footnote h would be deleted and the following 
corresponding changes would be made: 1) the reference to footnote h 
would be removed from Cf-252, 2) the entry for molybdenum-99 (Mo-99) 
would be revised to identify footnote h instead of footnote i, and 3) 
footnote i would be redesignated as footnote h. Footnote c in the entry 
for Ir-192 would be moved, so that it is clear that it applies only to 
iridium in special form. Footnote c would also be revised to 
specifically state that the activity of iridium in special form may be 
determined through measurement at a prescribed distance from the 
source. Table A-1 would be revised to include values for Kr-79. The 
A1 and A2 values for Kr-79 correspond to the 
A1 and A2 values in TS-R-1 (2009 edition) and the 
specific activity would be 4.2 x 10\4\ TBq/g (1.1 x 10\6\ Ci/g). The 
entry for Kr-81 would be revised to reflect that it is no longer the 
first entry for the isotopes of krypton. In addition, footnote a would 
be revised to identify the A1 and/or A2 values 
that include contributions from daughter radionuclides with half-lives 
of less than 10 days.
    Table A-2 would be revised to include values for Kr-79, reflect 
changes in TS-R-1 for the activity limit for exempt consignment for Te-
121m and in the list of parent radionuclides and their progeny included 
in secular equilibrium in Table A-2 in footnote b. The value for the 
activity concentration for exempt material for Kr-79 would be 1.0 x 
0\3\ Bq/g (2.7 x 10-8 Ci/g) and the value for the activity 
limit for exempt consignment would be 1.0 x 10\5\ Bq (2.7 x 
10-6 Ci). The activity limit for exempt consignment for Te-
121m would be revised from 1 x 10\5\ Bq (2.7 x 10-6 Ci) to 1 
x 10\6\ Bq (2.7 x 10-5 Ci). In footnote b, the chains for 
the parent radionuclides cerium-134 (Ce-134), Rn-220, Th-226, and U-240 
are proposed to be removed, and a chain for Ag-108m is proposed to be 
added. This would make footnote b to Table A-2 consistent with footnote 
b to Table 2 in TS-R-1. Changes in the list in footnote b were not 
initially made to TS-R-1 when the nuclide-specific basic radionuclide 
values from the International Basic Safety Standards (IAEA Safety 
Series No. 115, International Basic Safety Standards for Protection 
against Ionizing Radiation and for the Safety of Radiation Sources) 
were adopted for transportation purposes but were made in the 2005 
edition of TS-R-1.
    Table A-3 would be revised to reflect changes in TS-R-1. In the 
second entry, the descriptive phrase ``only alpha emitting 
radionuclides are known to be present'' would be changed to ``alpha 
emitting nuclides, but no neutron emitters, are known to be present'' 
to reduce the confusion caused by the current phrase, because all alpha 
emitting radionuclides also emit other particles and/or gamma rays. In 
the third entry, the descriptive phrase ``no relevant data are 
available'' would be changed to ``neutron emitting nuclides are known 
to be present or no relevant data are available'' to clarify that 
neutron-emitting radionuclides, or alpha emitters that also emit 
neutrons, such as Cf-252, Cf-254, and Cm-248, should be assigned to the 
third group. Footnote a would indicate the appropriate value of 
A1 for a group containing both alpha emitting radionuclides 
and beta or gamma emitting radionuclides when groups of radionuclides 
are based on the total alpha activity and the total beta and gamma 
activity.

V. Criminal Penalties

    For the purpose of Section 223 of the Atomic Energy Act (AEA), the 
Commission is proposing to amend 10 CFR part 71 under one or more of 
Sections 161b, 161i, or 161o of the AEA.

[[Page 29003]]

Willful violations of the rule would be subject to criminal 
enforcement.

VI. Agreement State Compatibility

    Under the ``Policy Statement on Adequacy and Compatibility of 
Agreement State Programs'' approved by the Commission on June 30, 1997, 
and published in the Federal Register (62 FR 46517; September 3, 1997), 
this rule would be a matter of compatibility between the NRC and the 
Agreement States, thereby providing consistency among the Agreement 
States' and the NRC requirements. The NRC staff analyzed the rule in 
accordance with the procedure established within part III, 
``Categorization Process for NRC Program Elements,'' of Handbook 5.9 to 
Management Directive 5.9, ``Adequacy and Compatibility of Agreement 
State Programs'' (ADAMS Accession No. ML041770094). The proposed 
compatibility categories assigned to the affected sections of 10 CFR 
part 71 are presented in the Compatibility Table in this section.
    There are four compatibility categories (A, B, C, and D). In 
addition, the NRC program elements can also be identified as having 
particular health and safety significance or as being reserved solely 
to the NRC. Compatibility Category A is assigned to 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 Compatibility 
Category A program elements in an essentially identical manner to 
provide uniformity in the regulation of agreement material on a 
nationwide basis. Compatibility Category B is assigned to those program 
elements that apply to activities that have direct and significant 
effects in multiple jurisdictions. An Agreement State should adopt 
Compatibility Category B program elements in an essentially identical 
manner. Compatibility Category C is assigned to those program elements 
that do not meet the criteria of Compatibility 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 Compatibility Category C program elements. 
Compatibility Category D is assigned to those program elements that do 
not meet any of the criteria of Compatibility Categories A, B, or C, 
and, thus, do not need to be adopted by Agreement States for purposes 
of compatibility.
    Health and Safety (H&S) are program elements that are not required 
for compatibility but are identified 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 H&S category based on those of 
the NRC that embody the essential objectives of the NRC program 
elements because of particular health and safety considerations. 
Compatibility Category NRC is assigned to those program elements that 
address areas of regulation that cannot be relinquished to Agreement 
States under the AEA, as amended, or provisions of 10 CFR. These 
program elements are not adopted by the Agreement States.
    The following table lists the parts and sections that would be 
revised and their corresponding categorization under the ``Policy 
Statement on Adequacy and Compatibility of Agreement State Programs.'' 
A bracket around a category means that the section may have been 
adopted elsewhere, and it is not necessary to adopt it again. The 
presence or absence of a bracket does not affect the compatibility 
category or the degree of uniformity required when an Agreement State 
adopts the requirement.

                                               Compatibility Table
----------------------------------------------------------------------------------------------------------------
                                                                                  Compatibility
       Section               Change              Subject       -------------------------------------------------
                                                                        Existing                  New\1\
----------------------------------------------------------------------------------------------------------------
71.0(d)(1)..........  Revised............  Purpose and Scope..  D......................  D.
71.4................  New................  Definition           .......................  [B].
                                            Contamination.
71.4................  Revised............  Definition           [B]....................  [B].
                                            Criticality Safety
                                            Index (CSI).
71.4................  Revised............  Definition Low       [B]....................  [B].
                                            Specific Activity
                                            (LSA) material.
71.4................  Revised............  Definition Special   [B]....................  [B].
                                            Form Radioactive
                                            Material.
71.4................  Revised............  Definition Uranium-- [B]....................  [B].
                                            natural, depleted,
                                            enriched.
71.6................  Revised............  Information          D......................  D.
                                            Collection
                                            Requirements: OMB
                                            Approval.
71.14(a)(1).........  Revised............  Exemption for low-   [B]....................  [B].
                                            level materials.
71.14(a)(2).........  Revised............  Exemption for low-   [B]....................  [B].
                                            level materials.
71.14(a)(3).........  New................  Exemption for low-   .......................  [B].
                                            level materials.
71.15(d)............  Revised............  Exemption from       [B]....................  [B].
                                            classification as
                                            fissile material.

[[Page 29004]]

 
71.17...............  Removal of brackets  General license:     [B]....................  B.
                       on Compatibility     NRC-approved
                       Category.            package.
71.17(c)............  Revised............  General license:     [B]....................  B.
                                            NRC-approved
                                            package.
71.19...............  Revised............  Previously approved  NRC....................  NRC.
                                            package.
71.21...............  Removal of brackets  General license:     [B]....................  B.
                       on Compatibility     Use of foreign
                       Category.            approved package.
71.21(a)............  Revised............  General license:     [B]....................  B.
                                            Use of foreign
                                            approved package.
71.21(d)............  Revised............  General license:     [B]....................  B.
                                            Use of foreign
                                            approved package.
71.31(b)............  Revised............  Contents of          NRC....................  NRC.
                                            application.
71.38...............  Retitled and         Renewal of a         NRC....................  NRC.
                       revised.             certificate of
                                            compliance.
71.70...............  New................  Incorporations by    .......................  NRC.
                                            reference.
71.75...............  Revised............  Qualification of     NRC....................  NRC.
                                            special form
                                            radioactive
                                            material.
71.85(a)............  Revised............  Preliminary          [B]....................  NRC.
                                            determinations.
71.85(b)............  Revised............  Preliminary          [B]....................  NRC.
                                            determinations.
71.85(c)............  Revised............  Preliminary          [B]....................  NRC.
                                            determinations.
71.85(d)............  New................  Preliminary          .......................  B.
                                            determinations.
71.91(a)............  Revised............  Records............  D......................  C.
71.91(b)............  Revised              Records............  D......................  NRC.
                       Compatibility
                       Category.
71.91(c)............  Revised              Records............  D......................  C.
                       Compatibility
                       Category.
71.91(d)............  Revised              Records............  D......................  C.
                       Compatibility
                       Category.
71.101(a)...........  Revised............  Quality assurance    D--For those States      C.
                                            requirements.        which have no users of  **Note: 10 CFR
                                                                 Type B packages--other   71.101(g) indicates
                                                                 than industrial          that QA programs for
                                                                 radiography.**.          industrial radiography
                                                                C--Those States which     Type B package users
                                                                 have users of Type B     are covered by Sec.
                                                                 packages--other than     34.31(b). It also
                                                                 industrial               indicated that this
                                                                 radiography**.           section satisfies Sec.
                                                                **Note: 10 CFR              71.17(b) and thus
                                                                 71.101(g) indicates      would satisfy those
                                                                 that QA programs for     sections referenced in
                                                                 industrial radiography   this provision (Sec.
                                                                 Type B package users     Sec.   71.101 through
                                                                 are covered by Sec.      71.137)
                                                                 34.31(b). It also
                                                                 indicated that this
                                                                 section satisfies Sec.
                                                                   71.12(b) and thus
                                                                 would satisfy those
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137)..
71.101(b)...........  Revised              Quality assurance    D--For those States      C.
                       Compatibility        requirements.        which have no users of  **Note: 10 CFR
                       Category.                                 Type B packages--other   71.101(g) indicates
                                                                 than industrial          that QA programs for
                                                                 radiography.**.          industrial radiography
                                                                C--Those States which     Type B package users
                                                                 have users of Type B     are covered by Sec.
                                                                 packages--other than     34.31(b). It also
                                                                 industrial               indicated that this
                                                                 radiography.**.          section satisfies Sec.
                                                                **Note: 10 CFR              71.17(b) and thus
                                                                 71.101(g) indicates      would satisfy those
                                                                 that QA programs for     sections referenced in
                                                                 industrial radiography   this provision (Sec.
                                                                 Type B package users     Sec.   71.101 through
                                                                 are covered by Sec.      71.137).
                                                                 34.31(b). It also
                                                                 indicated that this
                                                                 section satisfies Sec.
                                                                   71.12(b) and thus
                                                                 would satisfy those
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137).

[[Page 29005]]

 
71.101(c)(1)........  Revised              Quality assurance    D--For those States      C.
                       Compatibility        requirements.        which have no users of  **Note: 10 CFR
                       Category.                                 Type B packages--other   71.101(g) indicates
                                                                 than industrial          that QA programs for
                                                                 radiography**.           industrial radiography
                                                                C--Those States which     Type B package users
                                                                 have users of Type B     are covered by Sec.
                                                                 packages--other than     34.31(b). It also
                                                                 industrial               indicated that this
                                                                 radiography.**.          section satisfies Sec.
                                                                **Note: 10 CFR              71.17(b) and thus
                                                                 71.101(g) indicates      would satisfy those
                                                                 that QA programs for     sections referenced in
                                                                 industrial radiography   this provision (Sec.
                                                                 Type B package users     Sec.   71.101 through
                                                                 are covered by Sec.      71.137).
                                                                 34.31(b). It also
                                                                 indicated that this
                                                                 section satisfies Sec.
                                                                   71.12(b) and thus
                                                                 would satisfy those
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137).
71.101(c)(2)........  Revised............  Quality assurance    NRC....................  NRC.
                                            requirements.
71.101(g)...........  Revised              Quality assurance    C......................  C.
                       Compatibility        requirements.       **Note: 10 CFR           **Note: 10 CFR
                       Category Note.                            71.101(g) indicates      71.101(g) indicates
                                                                 that QA programs for     that QA programs for
                                                                 industrial radiography   industrial radiography
                                                                 Type B package users     Type B package users
                                                                 are covered by Sec.      are covered by Sec.
                                                                 34.31(b). It also        34.31(b). It also
                                                                 indicated that this      indicated that this
                                                                 section satisfies Sec.   section satisfies Sec.
                                                                   71.12(b) and thus        71.17(b) and thus
                                                                 would satisfy those      would satisfy those
                                                                 sections referenced in   sections referenced in
                                                                 this provision (Sec.     this provision (Sec.
                                                                 Sec.   71.101 through    Sec.   71.101 through
                                                                 71.137).                 71.137).
71.103(a)...........  Revised............  Quality assurance    D--For those States      C.
                                            organization.        which have no users of  **Note: Sec.
                                                                 Type B packages--other   71.101(g) indicates
                                                                 than industrial          that QA programs for
                                                                 radiography.**.          industrial radiography
                                                                [C]--Those States which   Type B package users
                                                                 have users of Type B     are covered by Sec.
                                                                 packages--other than     34.31(b). It also
                                                                 industrial               indicated that this
                                                                 radiography.**.          section satisfies Sec.
                                                                **Note: Sec.                71.17(b) and thus
                                                                 71.101(g) indicates      would satisfy those
                                                                 that QA programs for     sections referenced in
                                                                 industrial radiography   this provision (Sec.
                                                                 Type B package users     Sec.   71.101 through
                                                                 are covered by Sec.      71.137).
                                                                 34.31(b). It also
                                                                 indicated that this
                                                                 section satisfies Sec.
                                                                   71.12(b) and thus
                                                                 would satisfy those
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137).
71.103(b)...........  Revised              Quality assurance    C--Those States which    C.
                       Compatibility        organization.        have users of Type B    **Note: Sec.
                       Category Note.                            packages--other than     71.101(g) indicates
                                                                 industrial               that QA programs for
                                                                 radiography.**.          industrial radiography
                                                                **Note: Sec.              Type B package users
                                                                 71.101(g) indicates      are covered by Sec.
                                                                 that QA programs for     34.31(b). It also
                                                                 industrial radiography   indicated that this
                                                                 Type B package users     section satisfies Sec.
                                                                 are covered by Sec.        71.17(b) and thus
                                                                 34.31(b). It also        would satisfy those
                                                                 indicated that this      sections referenced in
                                                                 section satisfies Sec.   this provision (Sec.
                                                                   71.12(b) and thus      Sec.   71.101 through
                                                                 would satisfy those      71.137).
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137).
71.106..............  New................  Changes to quality   .......................  C.
                                            assurance program.
71.135..............  Revised............  Quality assurance    D--For those States      C.
                                            records.             which have no users of  **Note: 10 CFR
                                                                 Type B packages--other   71.101(g) indicates
                                                                 than industrial          that QA programs for
                                                                 radiography.**.          industrial radiography
                                                                C--For those States       Type B package users
                                                                 which have users of      are covered by Sec.
                                                                 Type B packages--other   34.31(b). It also
                                                                 than industrial          indicated that this
                                                                 radiography**.           section satisfies Sec.
                                                                **Note: 10 CFR              71.17(b) and thus
                                                                 71.101(g) indicates      would satisfy those
                                                                 that QA programs for     sections referenced in
                                                                 industrial radiography   this provision (Sec.
                                                                 Type B package users     Sec.   71.101 through
                                                                 are covered by Sec.      71.137).
                                                                 34.31(b). It also
                                                                 indicated that this
                                                                 section satisfies Sec.
                                                                   71.12(b) and thus
                                                                 would satisfy those
                                                                 sections referenced in
                                                                 this provision (Sec.
                                                                 Sec.   71.101 through
                                                                 71.137).

[[Page 29006]]

 
Appendix A..........  Revise paragraphs    Determination of A1  [B]....................  [B].
                       IV.a.--IV.f.;        and A2.
                       redesignate
                       paragraphs IV.c.--
                       IV.f. as
                       paragraphs IV.d.--
                       IV.g.; add
                       paragraph IV.c.;
                       redesignate the
                       text of paragraph
                       V. as paragraph
                       V.a.; and add
                       paragraph V.b.
Appendix A, Table A-  Revise entries for   A1 and A2 Values     [B]....................  [B].
 1.                    Cf-252, Ir-192, Kr-  for Radionuclides.
                       81, and Mo-99;
                       revise footnote a;
                       delete footnote h;
                       and redesignate
                       footnote i as
                       footnote h.
                      Add entry for Kr-
                       79..
Appendix A, Table A-  Add entry for Kr-    Exempt Material      [B]....................  [B].
 2.                    79; revise entries   Activity
                       for Kr-81 and Te-    Concentrations and
                       121m; and revise     Exempt Consignment
                       footnote b.          Activity Limits
                                            for Radionuclides.
Appendix A, Table A-  Revise entries for   General Values for   [B]....................  [B].
 3.                    column 1,            A1 and A2.
                       ``Contents,'' and
                       add footnote a.
----------------------------------------------------------------------------------------------------------------
\1\ Where there would be a change in the assigned compatibility category, a compatibility category is assigned,
  or the content of the section has been significantly changed, a summary of the analysis is presented in the
  following paragraphs. Changes in the assigned compatibility category are being made in Sec.  Sec.   71.4
  (added for the definition of contamination), 71.70, 71.85, 71.91, 71.101, 71.103, 71.106, and 71.135.

    In Sec.  71.4, the definition of contamination would be designated 
Compatibility Category [B], because it applies to activities that have 
direct and significant effects in multiple jurisdictions and it is also 
defined in the corresponding DOT regulations.
    In Sec. Sec.  71.17, 71.21, and 71.103, the compatibility category 
is unchanged, but the brackets were not retained because there are no 
corresponding DOT regulations.
    The new Sec.  71.70, ``Incorporations by reference,'' would be 
designated Compatibility Category NRC, because the documents 
incorporated by reference are incorporated for use in Sec.  71.75, 
which addresses activities under Federal jurisdiction.
    Section 71.85, ``Preliminary determinations,'' would be changed to 
make the requirements in Sec.  71.85(a) through (c) apply to holders of 
a CoC. Paragraphs 71.85(a) through (c) would be designated as 
Compatibility Category NRC, because they apply exclusively to 
certificate holders and the granting of the package approval is 
reserved to the NRC. Paragraph 71.85(d) would be added and applies to 
licensees. Paragraph 71.85(d) would be designated as Compatibility 
Category B because it applies to activities that have direct and 
significant effects in multiple jurisdictions and there is no 
corresponding DOT requirement.
    The compatibility category for Sec.  71.91, ``Records,'' would be 
changed from Compatibility Category D to Compatibility Category C. In 
reaching an agreement with the NRC, the States would have a general 
provision relating to records and for incident reporting. The 
recordkeeping requirements in Sec.  71.91 include requirements 
associated with transportation, which may involve multiple 
jurisdictions. With the exception of Sec.  71.91(b), the NRC is 
proposing to designate the compatibility of the requirements in Sec.  
71.91 as Compatibility Category C to require that the essential 
objectives of the requirements be adopted to avoid conflict, 
duplication, gaps, or other conditions that would jeopardize the 
orderly pattern in the regulation of agreement material on a nationwide 
basis, including creating an undue burden on interstate commerce 
through additional recordkeeping requirements; Sec.  71.91(b) only 
applies to CoC holders and applicants and would be designated as 
compatibility category NRC. The States would not be required to adopt 
them in an essentially identical manner as might be necessary if the 
requirements had a more direct and significant impact on multiple 
jurisdictions.
    In Sec.  71.101, the compatibility category would be simplified by 
removing the separate compatibility category for States that do not 
have a user of a Type B package. If a State does not have a user of a 
Type B package, the State is able to seek an exemption from the 
requirement to make their requirement compatible. The State 
requirements only need to be essentially compatible with respect to the 
requirements as they apply to licensees, because the application of the 
requirements to CoC holders and applicants would be performed by the 
NRC. The note that references the quality assurance programs for 
industrial radiographers would be updated by changing Sec.  71.12(b) to 
Sec.  71.17(b).
    In Sec.  71.103, the compatibility category for some users of 
packages was not

[[Page 29007]]

designated. The compatibility category would be simplified by removing 
the separate compatibility category for States that do not have a user 
of a Type B package and by removing the bracket around the 
compatibility category for Sec.  71.103(a). If a State does not have a 
user of a Type B package, the State would be able to seek an exemption 
from the requirement to make their requirement compatible. The State 
requirements only need to be essentially compatible with respect to the 
requirements as they apply to licensees, because the application of the 
requirements to CoC holders and applicants would be performed by the 
NRC. The note that references the quality assurance programs for 
industrial radiographers would be updated by changing Sec.  71.12(b) to 
Sec.  71.17(b).
    The new Sec.  71.106, ``Changes to quality assurance program,'' 
would apply to licensees and holders of, or applicants for, a CoC. The 
assigned compatibility category would be consistent with the other 
quality assurance requirements that apply to licensees. The State 
requirements only need to be essentially compatible with respect to the 
requirements as they apply to licensees, because the application of the 
requirements to CoC holders and applicants would be performed by the 
NRC.
    In Sec.  71.135, the compatibility category would be simplified by 
removing the separate compatibility category for States that do not 
have a user of a Type B package. If a State does not have a user of a 
Type B package, the State would be able to seek an exemption from the 
requirement to make their requirement compatible. The State 
requirements only need to be essentially compatible with respect to the 
requirements as they apply to licensees, because the application of the 
requirements to CoC holders and applicants would be performed by the 
NRC. The note that references the quality assurance programs for 
industrial radiographers would be updated by changing Sec.  71.12(b) to 
Sec.  71.17(b).

VII. Availability of Documents

    The following documents referenced in this proposed rulemaking are 
available either through ADAMS or at the NRC PDR:

----------------------------------------------------------------------------------------------------------------
               Document                         PDR                   ADAMS              ADAMS Accession No.
----------------------------------------------------------------------------------------------------------------
Management Directive 5.9, ``Adequacy   Yes..................  Yes..................  ML041770094
 and Compatibility of Agreement State
 Programs.''.
NRC Information Notice 2002-035:       Yes..................  Yes..................  ML023520339
 ``Changes to 10 CFR Parts 71 and 72
 Quality Assurance Programs.''
NRC Regulatory Issue Summary 2004-     Yes..................  Yes..................  ML042160293
 018: ``Expiration Date for 10 CFR
 Part 71 Quality Assurance Program
 Plan Approvals.''.
NUREG/CR-5342, ``Assessment and        Yes..................  Yes..................  ML12139A419
 Recommendations for Fissile-Material
 Packaging Exemptions and General
 Licenses within 10 CFR Part 71,''
 July 1998.
Draft Environmental Assessment and     Yes..................  Yes..................  ML12187A109
 Finding of No Significant Impact for
 the Proposed Rule Amending 10 CFR
 Part 71.
Draft Regulatory Analysis for          Yes..................  Yes..................  ML12187A110
 Proposed Rulemaking--Compatibility
 with IAEA Transportation Standards
 (10 CFR Part 71).
----------------------------------------------------------------------------------------------------------------

VIII. Plain Writing

    The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal 
agencies to write documents in a clear, concise, well-organized manner 
that also follows other best practices appropriate to the subject or 
field and the intended audience. The NRC has attempted to use plain 
language in promulgating this rule consistent with the Federal Plain 
Writing Act as well as the Presidential Memorandum, ``Plain Language in 
Government Writing,'' published June 10, 1998 (63 FR 31883). The NRC 
requests comments on the proposed rule with respect to the clarity and 
effectiveness of the language used. Comments should be sent to the NRC 
as explained in the ADDRESSES section of this document.

IX. Voluntary Consensus Standards

    The National Technology Transfer and Advancement Act of 1995 (Pub. 
L. 104-113) requires that Federal agencies use technical standards that 
are developed or adopted by voluntary consensus standards bodies unless 
the use of such a standard is inconsistent with applicable law or 
otherwise impractical. In this proposed rule, the NRC proposes to use 
and incorporate by reference the following consensus standards: 
International Organization for Standardization, ISO 2919:1999(E), 
``Radiation protection--Sealed radioactive sources--General 
requirements and classification,'' Second Edition (February 15, 1999), 
for the Class 4 and Class 5 impact tests and the Class 6 temperature 
test; and International Organization for Standardization, ISO 
9978:1992(E), ``Radiation protection--Sealed radioactive sources--
Leakage test methods,'' First Edition (February 15, 1992), for the 
leaktightness tests. The NRC invites comment on the applicability and 
use of other standards.
    In other portions of this proposed rule, the NRC is revising 
requirements that do not constitute the establishment of a standard 
that establishes generally applicable requirements. These revisions to 
the NRC requirements include changes to: (1) The scope of material 
falling under an existing exemption for natural materials and ores 
containing naturally occurring radionuclides at an activity 
concentration below a specified value; (2) conditions on general 
licenses; (3) the oversight of quality assurance programs, and (4) the 
removal of transitional arrangements for previously approved packages.

X. Finding of No Significant Environmental Impact: Availability

    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, not to prepare an environmental impact 
statement for this proposed rule because the Commission has concluded 
on the basis of an Environmental Assessment (ADAMS Accession No. 
ML12187A109) that this proposed rule, if adopted, would not be a major 
federal action significantly affecting the quality of the human 
environment.
    Many of the proposed changes fall under a categorical exclusion for 
which the Commission has previously determined that such actions, 
neither individually nor cumulatively, would have significant impacts 
on the human environment. The categorical exclusions in 10 CFR 
51.22(c)(2) and 10 CFR 51.22(c)(3) were used in the Environmental 
Assessment. The categorical exclusion at 10 CFR

[[Page 29008]]

51.22(c)(2) applies to amendments to 10 CFR part 71 that are corrective 
or of a minor or non-policy nature and do not substantially modify the 
regulations. The categorical exclusion at 10 CFR 51.22(c)(3) applies to 
amendments to 10 CFR part 71 that relate to: (i) Procedures for filing 
and reviewing applications for licenses or construction permit or early 
site permit or other forms of permission or for amendments to or 
renewals of licenses or construction permits or early site permits or 
other forms of permission; (ii) recordkeeping requirements; (iii) 
reporting requirements; (iv) education, training, experience, 
qualification, or other employment suitability requirements; or (v) 
actions on petitions for rulemaking relating to these amendments.
    Those changes not qualifying for a categorical exclusion were 
evaluated for their environmental impacts and include changes to: (1) 
Definitions; (2) the exemption of low-level materials; (3) the fissile 
material exemption for low-enriched fissile material; (4) alternate 
tests that may be used for the qualification of special form material; 
(5) preliminary determinations; (6) the A1 and A2 
values for radionuclides; and (7) the exempt material activity 
concentrations and exempt consignment activity limits for 
radionuclides. The effects of these changes are addressed in more 
detail in the Environmental Assessment. The changes to the fissile 
material exemption would further reduce the potential for criticality 
during the transport of low-enriched fissile material under the fissile 
material exemption. Other changes, such as those relating to the 
exemption of low-level material, the A1 and A2 
values for radionuclides, and the exempt material activity 
concentrations and exempt consignment activity limits for radionuclides 
have been found to have small or very small impacts. Some natural 
material and ore may be shipped without being regulated as hazardous 
material. The low-level material exemption would be changed to allow 
some additional material to be transported without being regulated as 
hazardous material. The amount of transported material affected by this 
change is a very small fraction of the material that already qualifies 
for the exemption and would be allowed no greater activity than is 
already allowed for material that may already be transported under the 
exemption. Although there are changes to A1 and 
A2 values--used to determine the type of packaging, the 
exempt material activity concentrations, and the exempt consignment 
activity limits for some radionuclides, the approach for determining 
the appropriate values has not changed, so there would be very small 
impacts from these changes.
    The determination of this Environmental Assessment is that there 
will be no significant impact to the public from this action. However, 
the NRC is providing an opportunity to comment on the Environmental 
Assessment. Comments on any aspect of the Environmental Assessment may 
be submitted to the NRC as indicated under the ADDRESSES section of 
this document.
    The NRC has sent a copy of the Environmental Assessment and this 
proposed rule to every State Liaison Officer and requested their 
comments on the Environmental Assessment.

XI. Paperwork Reduction Act Statement

    This proposed rule contains new or amended information collection 
requirements that are subject to the Paperwork Reduction Act of 1995 
(44 U.S.C. 3501 et seq). This proposed rule has been submitted to the 
Office of Management and Budget (OMB) for review and approval of the 
information collection requirements.
    Type of submission, new or revision: Revision.
    The title of the information collection: 10 CFR part 71, Revisions 
to Transportation Safety Requirements and Harmonization with 
International Atomic Energy Agency Transportation Requirements.
    The form number if applicable: Not applicable.
    How often the collection is required: On occasion, for reports of 
changes reducing commitments to the NRC on quality assurance plans. 
Every 24 months for all changes to quality assurance plans.
    Who will be required or asked to report: General licensees or users 
of packages, certificate holders and certificate applicants.
    An estimate of the number of annual responses: 31.
    The estimated number of annual respondents: 250.
    An estimate of the total number of hours needed annually to 
complete the requirement or request: -1,700 hours (a decrease of 1,925 
hours reporting + an increase of 100 third party disclosure hours and 
125 hours recordkeeping).
    Abstract: The NRC is proposing to amend its regulations for the 
packaging and transportation of radioactive material, including changes 
to information collections that would affect persons with a quality 
assurance program approved under 10 CFR part 71. Rather than submitting 
all quality assurance program changes to the NRC for approval, 
licensees, certificate holders, and applicants would only need to 
submit changes to their quality assurance program that would reduce 
their commitments to the NRC. They would be required to keep records of 
all quality assurance program changes and submit a report of these 
changes to the NRC every 24 months. Burden on licensees would be 
reduced for renewing quality assurance programs, as future approvals of 
these programs would not expire.
    The NRC is seeking public comment on the potential impact of the 
information collections contained in this proposed rule (or proposed 
policy statement) 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?
    A copy of the OMB clearance package may be viewed free of charge at 
the NRC PDR, Room O1-F21, One White Flint North, 11555 Rockville Pike, 
Rockville, MD 20852. The OMB clearance package and rule are available 
at the NRC public Web site: http://www.nrc.gov/public-involve/doc-comment/omb/index.html, for 60 days after the signature date of this 
document.
    Send comments on any aspect of these proposed information 
collections, including suggestions for reducing the burden and on the 
previously stated issues, by June 17, 2013 to the Information Services 
Branch (T-5 F53), U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, or by email to [email protected] and to the 
Desk Officer, Chad Whiteman, Office of Information and Regulatory 
Affairs, NEOB-10202, (3150-0008), Office of Management and Budget, 
Washington, DC 20503. Comments on the proposed information collections 
may also be submitted via the Federal rulemaking Web site http://www.regulations.gov, docket # NRC-2008-0198. 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. 
Comments can also be emailed to [email protected] or

[[Page 29009]]

submitted by telephone at 202-395-4718.

Public Protection Notification

    The NRC may not conduct or sponsor, and a person is not required to 
respond to, a request for information or an information collection 
requirement unless the requesting document displays a currently valid 
OMB control number.

XII. Regulatory Analysis

    The Commission has prepared a draft regulatory analysis (ADAMS 
Accession No. ML12187A110) on this proposed regulation. The analysis 
examines the costs and benefits of the alternatives considered by the 
Commission. The Commission requests public comment on the draft 
regulatory analysis. Comments on the draft analysis may be submitted to 
the NRC as indicated under the ADDRESSES section of this document.

XIII. Regulatory Flexibility Certification

    In accordance with the Regulatory Flexibility Act of 1980 (5 U.S.C. 
605(b)), the Commission certifies that this rule would not, if 
promulgated, have a significant economic impact on a substantial number 
of small entities. This rule affects NRC licensees who transport or 
deliver to a carrier for transport, relatively large quantities of 
radioactive material in a single package; holders of a quality 
assurance program description issued under 10 CFR parts 50, 71, or 72; 
and holders of a certificate of compliance for a transportation 
package. These companies do not typically 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 at 10 CFR 
2.810. Also, a draft regulatory analysis was performed for this 
proposed rule. The regulatory analysis included an evaluation of the 
costs associated with the proposed requirements. The proposed 
rulemaking includes changes that would reduce the regulatory burden for 
licensees and certificate holders. Based on the information developed 
in the regulatory analysis, it is believed that there will not be 
significant economic impacts on a substantial number of small entities.

XIV. Backfitting

    The NRC has determined that the backfit rule (50.109, 70.76, 72.62, 
or 76.76) and the issue finality provisions in 10 CFR part 52 do not 
apply to this proposed rule because this amendment would not involve 
any provisions that would impose backfits as defined in 10 CFR Chapter 
I. Therefore, a backfit analysis is not required for this proposed 
rule, and the NRC did not prepare a backfit analysis for this proposed 
rule.

List of Subjects in 10 CFR Part 71

    Criminal penalties, Hazardous materials transportation, Nuclear 
materials, 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 NRC is proposing to 
adopt the following amendments to 10 CFR part 71:

PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

0
1. The authority citation for part 71 continues to read as follows:

    Authority: Atomic Energy Act secs. 53, 57, 62, 63, 81, 161, 182, 
183, 223, 234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 
2232, 2233, 2273, 2282, 2297f); Energy Reorganization Act secs. 201, 
202, 206, 211 (42 U.S.C. 5841, 5842, 5846, 5851); Nuclear Waste 
Policy Act sec. 180 (42 U.S.C. 10175); Government Paperwork 
Elimination Act sec. 1704 (44 U.S.C. 3504 note); Energy Policy Act 
of 2005, Pub. L. No. 109-58, 119 Stat. 594 (2005). Section 71.97 
also issued under sec. 301, Pub. L. 96-295, 94 Stat. 789-790.


Sec.  71.0  [Amended]

0
2. In Sec.  71.0, paragraph (d)(1), remove the reference to 
``Sec. Sec.  71.20 through 72.23'' and add, in its place, the reference 
``Sec. Sec.  71.21 through 71.23''.
0
3. In Sec.  71.4, add in alphabetical order the definition of 
``contamination,'' and revise the definitions of ``Criticality Safety 
Index (CSI),'' ``Low Specific Activity (LSA) material,'' ``Special form 
radioactive material,'' and ``Uranium--natural, depleted, enriched'' to 
read as follows:


Sec.  71.4  Definitions.

* * * * *
    Contamination means the presence of a radioactive substance on a 
surface in quantities in excess of 0.4 Bq/cm\2\ for beta and gamma 
emitters and low toxicity alpha emitters, or 0.04 Bq/cm\2\ for all 
other alpha emitters.
    (1) Fixed contamination means contamination that cannot be removed 
from a surface during normal conditions of transport.
    (2) Non-fixed contamination means contamination that can be removed 
from a surface during normal conditions of transport.
* * * * *
    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, overpacks or freight containers containing 
fissile material during transportation. Determination of the 
criticality safety index is described in Sec. Sec.  71.22, 71.23, and 
71.59 of this part. The criticality safety index for an overpack, 
freight container, consignment or conveyance containing fissile 
material packages is the arithmetic sum of the criticality safety 
indices of all the fissile material packages contained within the 
overpack, freight container, consignment or conveyance.
* * * * *
    Low Specific Activity (LSA) material means radioactive material 
with limited specific activity which is nonfissile or is excepted under 
Sec.  71.15 of this part, and which satisfies the descriptions and 
limits set forth below. Shielding materials surrounding the LSA 
material may not be considered in determining the estimated average 
specific activity of the package contents. The LSA material must be in 
one of three groups:
    (1) LSA-I.
    (i) Uranium and thorium ores, concentrates of uranium and thorium 
ores, and other ores containing naturally occurring radionuclides that 
are intended to be processed for the use of these radionuclides;
    (ii) Natural uranium, depleted uranium, natural thorium or their 
compounds or mixtures, provided they are unirradiated and in solid or 
liquid form;
    (iii) Radioactive material other than fissile material, for which 
the A2 value is unlimited; or
    (iv) Other radioactive material in which the activity is 
distributed throughout and the estimated average specific activity does 
not exceed 30 times the value for exempt material activity 
concentration determined in accordance with appendix A.
    (2) LSA-II.
    (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
    (ii) Other material in which the activity 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), excluding powders, that satisfy the requirements of Sec.  
71.77 of this part, in which:
    (i) The radioactive material is distributed throughout a solid or a 
collection of solid objects, or is

[[Page 29010]]

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 estimated average specific activity of the solid, 
excluding any shielding material, does not exceed 2 x 10-3 
A2/g.
* * * * *
    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 of this part. A 
special form encapsulation designed in accordance with the requirements 
of Sec.  71.4 of this part in effect on June 30, 1983 (see 10 CFR part 
71, revised as of January 1, 1983), and constructed before July 1, 
1985; a special form encapsulation designed in accordance with the 
requirements of Sec.  71.4 of this part in effect on March 31, 1996 
(see 10 CFR part 71, revised as of January 1, 1996), and constructed 
before April 1, 1998; and special form material that was successfully 
tested before [EFFECTIVE DATE OF FINAL RULE] in accordance with the 
requirements of Sec.  71.75(d) of this part in effect before [EFFECTIVE 
DATE OF FINAL RULE] may continue to be used. Any other special form 
encapsulation must meet the specifications of this definition.
* * * * *
    Uranium--natural, depleted, enriched:
    (1) Natural uranium means uranium (which may be chemically 
separated) 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.
0
4. In Sec.  71.6, paragraph (b) is revised to read as follows:


Sec.  71.6  Information collection requirements: OMB approval.

* * * * *
    (b) The approved information collection requirements contained in 
this part appear in Sec. Sec.  71.5, 71.7, 71.9, 71.12, 71.17, 71.19, 
71.22, 71.23, 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, 71.106, 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, 
and appendix A, paragraph II.
0
5. In Sec.  71.14, paragraphs (a)(1) and (2) are revised,and paragraph 
(a)(3) is added to read as follows:


Sec.  71.14  Exemption for low-level materials.

    (a) * * *
    (1) Natural material and ores containing naturally occurring 
radionuclides that are either in their natural state, or have only been 
processed for purposes other than for the extraction of the 
radionuclides, and which are not intended to be processed for the use 
of these radionuclides, provided the activity concentration of the 
material does not exceed 10 times the applicable radionuclide activity 
concentration values specified in appendix A, Table A-2, or Table A-3, 
of this part.
    (2) Materials for which the activity concentration is not greater 
than the activity concentration values specified in appendix A, Table 
A-2, or Table A-3 of this part, or for which the consignment activity 
is not greater than the limit for an exempt consignment found in 
appendix A, Table A-2, or Table A-3, of this part.
    (3) Non-radioactive solid objects with radioactive substances 
present on any surfaces in quantities not in excess of the levels cited 
in the definition of contamination in Sec.  71.4 of this part.
* * * * *
0
6. In Sec.  71.15, paragraph (d) is revised to read as follows:


Sec.  71.15  Exemption from classification as fissile material.

* * * * *
    (d) 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 
constitutes less than 5 percent of the uranium mass, and that the 
fissile material is distributed homogeneously and does not form a 
lattice arrangement within the package.
* * * * *
0
7. In Sec.  71.17, paragraph (c) introductory text, (c)(1), and (c)(2) 
are revised to read as follows:


Sec.  71.17  General license: NRC-approved package.

* * * * *
    (c) Each licensee issued a general license under paragraph (a) of 
this section shall--
    (1) Maintain a copy of the CoC, or other approval of the package, 
and 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) Comply 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
* * * * *
0
8. In Sec.  71.19, paragraphs (b) through (e) are redesignated as 
paragraphs (a) through (d), and redesignated paragraph (b)(2) is 
revised to read as follows:


Sec.  71.19  Previously approved package.

* * * * *
    (b) * * *
    (2) A package used for a shipment to a location outside the United 
States is subject to multilateral approval as defined in the DOT 
regulations at 49 CFR 173.403.
* * * * *
0
9. In Sec.  71.21, paragraphs (a) and (d) are revised to read as 
follows:


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 
171.23.
* * * * *
    (d) Each licensee issued a general license under paragraph (a) of 
this section shall--
    (1) Maintain 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) Comply with the terms and conditions of the certificate and 
revalidation, and with the applicable requirements of subparts A, G, 
and H of this part.

[[Page 29011]]

Sec.  71.31  [Amended]

0
1. In Sec.  71.31, paragraph (b), remove the reference to ``Sec.  
71.13'' and add, in its place, the reference to ``Sec.  71.19''.
0
2. Section 71.38 is revised to read as follows:


Sec.  71.38  Renewal of a certificate of compliance.

    (a) Except as provided in paragraph (b) of this section, each 
Certificate of Compliance expires at the end of the day, in the month 
and year stated in the approval.
    (b) In any case in which a person, not less than 30 days before the 
expiration of an existing Certificate of Compliance issued pursuant to 
the part, has filed an application in proper form for renewal, the 
existing Certificate of Compliance for which the renewal application 
was filed shall not be deemed to have expired until final action on the 
application for renewal has been taken by the Commission.
    (c) In applying for renewal of an existing Certificate of 
Compliance, an applicant may be required to submit a consolidated 
application that is comprised of as few documents as possible. The 
consolidated application should incorporate all changes to its 
certificate, including changes that are incorporated by reference in 
the existing certificate.
0
3. Add Sec.  71.70 to subpart F to read as follows:


Sec.  71.70  Incorporations by reference.

    (a) The materials listed in this section are incorporated by 
reference in the corresponding sections noted and made a part of the 
regulations in 10 CFR part 71. These incorporations by reference were 
approved by the Director of the Federal Register under 5 U.S.C. 552(a) 
and 1 CFR part 51. These materials are incorporated as they exist on 
the date of the approval. A notice of any changes made to the material 
incorporated by reference will be published in the Federal Register and 
the material must be available to the public. The materials are 
available for purchase at the corresponding address noted in this 
section. The materials can also be examined at the NRC Public Document 
Room, O1-F21, 11555 Rockville Pike, Rockville, Maryland 20852 or at the 
NRC Library, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland 20852; telephone: 301-415-5610; email: 
[email protected]. The materials are also available for 
inspection at the National Archives and Records Administration (NARA). 
For information on the availability of this material at NARA, call 202-
741-6030, or go to http://www.archives.gov/federal-register/cfr/ibr-locations.html.
    (b) The following material is available for purchase from the 
American National Standards Institute, 25 West 43rd Street, 4th floor, 
New York, NY 10036, 212-642-4900, http://www.ansi.org, or 
[email protected].
    (1) International Organization for Standardization, ISO 
9978:1992(E), ``Radiation protection--Sealed radioactive sources--
Leakage test methods,'' First Edition (February 15, 1992), 
incorporation by reference approved for Sec.  71.75(a) of this part.
    (2) International Organization for Standardization, ISO 
2919:1999(E), ``Radiation protection--Sealed radioactive sources--
General requirements and classification,'' Second Edition (February 15, 
1999), incorporation by reference approved for Sec.  71.75(d) of this 
part.
0
4. In Sec.  71.75, paragraphs (a)(5), (b)(2)(ii), (b)(2)(iii), (d)(1), 
and (d)(2) are revised to read as follows:


Sec.  71.75  Qualification of special form radioactive material.

    (a) * * *
    (5) A specimen that comprises or simulates radioactive material 
contained in a sealed capsule need not be subjected to the 
leaktightness procedure specified in this section, provided it is 
alternatively subjected to any of the tests prescribed in ISO 
9978:1992(E), ``Radiation protection--Sealed radioactive sources--
Leakage test methods'' (incorporated by reference in Sec.  71.70 of 
this part).
    (b) * * *
    (2) * * *
    (ii) The flat face of the billet must be 25 millimeters (mm) (1 
inch) in diameter with the edge rounded off to a radius of 3 mm  0.3 mm (.12 in  0.012 in);
    (iii) The lead must be hardness number 3.5 to 4.5 on the Vickers 
scale and not more than 25 mm (1 inch) thick, and must cover an area 
greater than that covered by the specimen;
* * * * *
    (d) * * *
    (1) The impact test and the percussion test of this section, 
provided that the specimen is:
    (i) Less than 200 grams and alternatively subjected to the Class 4 
impact test prescribed in ISO 2919:1999(E), ``Radiation protection--
Sealed radioactive sources--General requirements and classification'' 
(incorporated by reference in Sec.  71.70 of this part); or
    (ii) Less than 500 grams and alternatively subjected to the Class 5 
impact test prescribed in ISO 2919:1999(E), ``Radioactive protection--
Sealed radioactive sources--General requirements and classification'' 
(incorporated by reference in Sec.  71.70 of this part); and
    (2) The heat test of this section, provided the specimen is 
alternatively subjected to the Class 6 temperature test specified in 
ISO 2919:1999(E), ``Radioactive protection--Sealed radioactive 
sources--General requirements and classification'' (incorporated by 
reference in Sec.  71.70 of this part).
0
5. In Sec.  71.85, paragraphs (a), (b), and (c) are revised and 
paragraph (d) is added to read as follows:


Sec.  71.85  Preliminary determinations.

* * * * *
    (a) The certificate holder shall ascertain that there are no 
cracks, pinholes, uncontrolled voids, or other defects that could 
significantly reduce the effectiveness of the packaging;
    (b) Where the maximum normal operating pressure will exceed 35 kPa 
(5 lbf/in\2\) gauge, the certificate holder shall test the containment 
system at an internal pressure at least 50 percent higher than the 
maximum normal operating pressure, to verify the capability of that 
system to maintain its structural integrity at that pressure;
    (c) The certificate holder shall conspicuously and durably mark the 
packaging with its model number, serial number, gross weight, and a 
package identification number assigned by the NRC. Before applying the 
model number, the certificate holder shall determine that the packaging 
has been fabricated in accordance with the design approved by the 
Commission; and
    (d) The licensee shall ascertain that the determinations in 
paragraphs (a) through (c) of this section have been made.


Sec.  71.91  [Amended]

0
1. In Sec.  71.91, paragraph (a), remove the reference to ``Sec.  
71.10'' and add, in its place, the reference to ``Sec.  71.14''.
0
2. In Sec.  71.101, paragraphs (a) and (c)(2) are revised to read as 
follows:


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

[[Page 29012]]

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. Each certificate 
holder and applicant for a package approval is responsible for 
satisfying the quality assurance requirements that apply to design, 
fabrication, testing, and modification of packaging subject to this 
subpart. Each licensee is responsible for satisfying the quality 
assurance requirements that apply to its use of a packaging for the 
shipment of licensed material subject to this subpart.
* * * * *
    (c) * * *
    (2) Before the fabrication, testing, or modification of any package 
for the shipment of licensed material subject to this subpart, each 
certificate holder, or applicant for a Certificate of Compliance (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 of this part, 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.
* * * * *
0
3. In Sec.  71.103, paragraph (a) is revised to read as follows:


Sec.  71.103  Quality assurance organization.

    (a) The licensee, certificate holder, and applicant for a 
Certificate of Compliance (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. These activities include performing the functions 
associated with attaining quality objectives and the quality assurance 
functions.
* * * * *
0
4. Add Sec.  71.106 to subpart H to read as follows:


Sec.  71.106  Changes to quality assurance program.

    (a) Each quality assurance program approval holder shall submit, in 
accordance with Sec.  71.1(a) of this part, a description of a proposed 
change to its NRC-approved quality assurance program that would reduce 
commitments in the program description as approved by the NRC. The 
quality assurance program approval holder shall not implement the 
change before receiving NRC approval.
    (1) The description of a proposed change to the NRC-approved 
quality assurance program must identify the change, the reason for the 
change, and the basis for concluding that the revised program 
incorporating the change continues to satisfy the applicable 
requirements of subpart H of this part.
    (2) [Reserved]
    (b) Each quality assurance program approval holder may change a 
previously approved quality assurance program without prior NRC 
approval, if the change does not reduce the commitments in the quality 
assurance program previously approved by the NRC. Changes to the 
quality assurance program that do not reduce the commitments shall be 
submitted to the NRC every 24 months, in accordance with Sec.  71.1(a) 
of this part. In addition to quality assurance program changes 
involving administrative improvements and clarifications; spelling 
corrections; and non-substantive changes to punctuation or editorial 
items; the following changes are not considered reductions in 
commitment:
    (1) The use of a quality assurance standard approved by the NRC 
that is more recent than the quality assurance standard in the 
certificate holder's or applicant's current quality assurance program 
at the time of the change;
    (2) The use of generic organizational position titles that clearly 
denote the position function, supplemented as necessary by descriptive 
text, rather than specific titles, provided that there is no 
substantive change to either the functions of the position or reporting 
responsibilities;
    (3) The use of generic organizational charts to indicate functional 
relationships, authorities, and responsibilities, or alternatively, the 
use of descriptive text, provided that there is no substantive change 
to the functional relationships, authorities, or responsibilities;
    (4) The elimination of quality assurance program information that 
duplicates language in quality assurance regulatory guides and quality 
assurance standards to which the quality assurance program approval 
holder has committed to on record; and
    (5) Organizational revisions that ensure that persons and 
organizations performing quality assurance functions continue to have 
the requisite authority and organizational freedom, including 
sufficient independence from cost and schedule when opposed to safety 
considerations.
    (c) Each quality assurance program approval holder shall maintain 
records of quality assurance program changes.
0
5. Section 71.135 is revised to read as follows:


Sec.  71.135  Quality assurance records.

    The licensee, certificate holder, and applicant for a Certificate 
of Compliance (CoC) shall maintain sufficient written records to 
describe the activities affecting quality. These records must include 
changes to the quality assurance program as required by Sec.  71.106 of 
this part, the instructions, procedures, and drawings required by Sec.  
71.111 of this part to prescribe quality assurance activities and 
closely related specifications such as required qualifications of 
personnel, procedures, and equipment. The records must include the 
instructions or procedures that 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 quality 
assurance program, 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.
0
6. In appendix A to part 71, IV.a. and IV.b. are revised, paragraphs 
IV.c. through IV.f. are redesignated as paragraphs IV.d. through IV.g. 
and are revised, new paragraph IV.c. is added, paragraph V. is 
redesignated as paragraph V.a., and new paragraph V.b. is added to read 
as follows:

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

* * * * *
    IV. * * *
    a. For special form radioactive material, the maximum quantity 
transported in a Type A package is as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.010

Where B(i) is the activity of radionuclide i in special form, 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] TP16MY13.011


[[Page 29013]]


Where B(i) is the activity of radionuclide i in normal form, and 
A2(i) is the A2 value for radionuclide i.
    c. If the package contains both special and normal form 
radioactive material, the activity that may be transported in a Type 
A package is as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.012

Where B(i) is the activity of radionuclide i as special form 
radioactive material, A1(i) is the A1 value 
for radionuclide i, C(j) is the activity of radionuclide j as normal 
form radioactive material, and A2(j) is the A2 
value for radionuclide j.
    d. Alternatively, the A1 value for mixtures of 
special form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.013

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.
    e. Alternatively, the A2 value for mixtures of normal 
form material may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.014

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.
    f. The exempt activity concentration for mixtures of nuclides 
may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.015

Where f(i) is the fraction of activity concentration of radionuclide 
i in the mixture and [A](i) is the activity concentration for exempt 
material containing radionuclide i.
    g. The activity limit for an exempt consignment for mixtures of 
radionuclides may be determined as follows:
[GRAPHIC] [TIFF OMITTED] TP16MY13.016

Where f(i) is the fraction of activity of radionuclide i in the 
mixture and A(i) is the activity limit for exempt consignments for 
radionuclide i.
    V.a. * * *
    b. 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 [A] (activity 
concentration for exempt material) or A (activity limit for exempt 
consignment) value, as appropriate, for the radionuclides in each 
group may be used in applying the formulas in paragraph IV of this 
appendix. Groups may be based on the total alpha activity and the 
total beta/gamma activity when these are known, using the lowest [A] 
or A values for the alpha emitters and beta/gamma emitters, 
respectively.
* * * * *
0
7. In appendix A to part 71, Table A-1:
0
a. Add entry for Kr-79 in alphanumeric order;
0
b. Revise the entries for Cf-252, Ir-192, Kr-81, and Mo-99;
0
c. Revise footnotes a and c; and
0
d. Remove footnote h; and
0
e. Redesignate footnote i as footnote h.
    The revisions read as follows:

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

* * * * *

--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                                                                 Specific activity
     Symbol of radionuclide       Element and atomic No.     A1 (TBq)       A1 (Ci)\b\       A2 (TBq)       A2 (Ci)\b\   -------------------------------
                                                                                                                              (TBq/g)         (Ci/g)
--------------------------------------------------------------------------------------------------------------------------------------------------------
 
                                                                      * * * * * * *
Cf-252.........................  .......................    1.0 x 10-\1\             2.7    3.0 x 10-\3\    8.1 x 10-\2\     2.0 x 10\1\     5.4 x 10\2\
 
                                                                      * * * * * * *
Ir-192.........................  .......................         \c\ 1.0    \c\ 2.7 x 10    6.0 x 10-\1\     1.6 x 10\1\     3.4 x 10\2\     9.2 x 10\3\
 
                                                                      * * * * * * *
Kr-79..........................  Krypton (36)...........             4.0     1.1 x 10\2\             2.0     5.4 x 10\1\     4.2 x 10\4\     1.1 x 10\6\
Kr-81..........................  .......................     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\
 
                                                                      * * * * * * *
Mo-99 (a)(h)...................  .......................             1.0     2.7 x 10\1\    6.0 x 10-\1\     1.6 x 10\1\     1.8 x 10\4\     4.8 x 10\5\
 
                                                                      * * * * * * *
--------------------------------------------------------------------------------------------------------------------------------------------------------
\a\ A1 and/or A2 values include contributions from daughter nuclides with half-lives less than 10 days, as listed in the following:


 
 
 
Mg-28                   Al-28
Ca-47                   Sc-47
Ti-44                   Sc-44
Fe-52                   Mn-52m
Fe-60                   Co-60m
Zn-69m                  Zn-69

[[Page 29014]]

 
Ge-68                   Ga-68
Rb-83                   Kr-83m
Sr-82                   Rb-82
Sr-90                   Y-90
Sr-91                   Y-91m
Sr-92                   Y-92
Y-87                    Sr-87m
Zr-95                   Nb-95m
Zr-97                   Nb-97m, Nb-97
Mo-99                   Tc-99m
Tc-95m                  Tc-95
Tc-96m                  Tc-96
Ru-103                  Rh-103m
Ru-106                  Rh-106
Pd-103                  Rh-103m
Ag-108m                 Ag-108
Ag-110m                 Ag-110
Cd-115                  In-115m
In-114m                 In-114
Sn-113                  In-113m
Sn-121m                 Sn-121
Sn-126                  Sb-126m
Te-127m                 Te-127
Te-129m                 Te-129
Te-131m                 Te-131
Te-132                  I-132
I-135                   Xe-135m
Xe-122                  I-122
Cs-137                  Ba-137m
Ba-131                  Cs-131
Ba-140                  La-140
Ce-144                  Pr-144m, Pr-144
Pm-148m                 Pm-148
Gd-146                  Eu-146
Dy-166                  Ho-166
Hf-172                  Lu-172
W-178                   Ta-178
W-188                   Re-188
Re-189                  Os-189m
Os-194                  Ir-194
Ir-189                  Os-189m
Pt-188                  Ir-188
Hg-194                  Au-194
Hg-195m                 Hg-195
Pb-210                  Bi-210
Pb-212                  Bi-212, Tl-208, Po-212
Bi-210m                 Tl-206
Bi-212                  Tl-208, Po-212
At-211                  Po-211
Rn-222                  Po-218, Pb-214, At-218, Bi-214, Po-214
Ra-223                  Rn-219, Po-215, Pb-211, Bi-211, Po-211, Tl-207
Ra-224                  Rn-220, Po-216, Pb-212, Bi-212, Tl-208, Po-212
Ra-225                  Ac-225, Fr-221, At-217, Bi-213, Tl-209, Po-213,
                         Pb-209
Ra-226                  Rn-222, Po-218, Pb-214, At-218, Bi-214, Po-214
Ra-228                  Ac-228
Ac-225                  Fr-221, At-217, Bi-213, Tl-209, Po-213, Pb-209
Ac-227                  Fr-223
Th-228                  Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208,
                         Po-212
Th-234                  Pa-234m, Pa-234
Pa-230                  Ac-226, Th-226, Fr-222, Ra-222, Rn-218, Po-214
U-230                   Th-226, Ra-222, Rn-218, Po-214
U-235                   Th-231
Pu-241                  U-237
Pu-244                  U-240, Np-240m
Am-242m                 Am-242, Np-238
Am-243                  Np-239
Cm-247                  Pu-243
Bk-249                  Am-245
Cf-253                  Cm-249
 
                              * * * * * * *
 
\c\ The activity of Ir-192 in special form 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.
 * * * * * * *
\h\ A2 = 0.74 TBq (20 Ci) for Mo-99 for domestic use.


[[Page 29015]]

* * * * *
0
8. In appendix A, Table A-2, the entry for Kr-79 is added, in 
alphanumeric order, the entries for Kr-81 and Te-121m are revised, and 
footnote b is revised to read as follows:

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

* * * * *

                       Table A-2--Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                   Activity           Activity
                                                              Element and     concentration for  concentration for    Activity limit     Activity limit
                  Symbol of radionuclide                       atomic No.      exempt material    exempt material       for exempt         for exempt
                                                                                    (Bq/g)             (Ci/g)        consignment (Bq)   consignment (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
 
                                                                      * * * * * * *
Kr-79....................................................       Krypton (36)        1.0 x 10\3\       2.7 x 10-\8\        1.0 x 10\5\       2.7 x 10-\6\
Kr-81....................................................  .................        1.0 x 10\4\       2.7 x 10-\7\        1.0 x 10\7\       2.7 x 10-\4\
 
                                                                      * * * * * * *
Te-121m..................................................  .................        1.0 x 10\2\       2.7 x 10-\9\        1.0 x 10\6\       2.7 x 10-\5\
 
                                                                      * * * * * * *
--------------------------------------------------------------------------------------------------------------------------------------------------------
 * * * * * * *
\b\ Parent nuclides and their progeny included in secular equilibrium are listed as follows:


 
 
 
Sr-90                   Y-90
Zr-93                   Nb-93m
Zr-97                   Nb-97
Ru-106                  Rh-106
Ag-108m                 Ag-108
Cs-137                  Ba-137m
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-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-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, Pb-210, Bi-210,
                         Po-210
Np-237                  Pa-233
Am-242m                 Am-242
Am-243                  Np-239
 

* * * * *
0
9. In appendix A to part 71, Table A-3 is revised to read as follows:

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

* * * * *

                                                         Table A-3--General Values for A1 and A2
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                A1                      A2              Activity     Activity
                                                     ------------------------------------------------   concen-      concen-      Activity     Activity
                                                                                                      tration for  tration for   limits for   limits for
                      Contents                                                                           exempt       exempt       exempt       exempt
                                                         (TBq)       (Ci)        (TBq)       (Ci)       material     material     consign-     consign-
                                                                                                         (Bq/g)       (Ci/g)     ments (Ba)   ments (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
Only beta or gamma emitting radionuclides are known   1 x 10-\1\       2.7 x    2 x 10 -   5.4 x 10-    1 x 10\1\    2.7 x 10-    1 x 10\4\    2.7 x 10-
 to be present......................................                   10\0\         \2\         \1\                      \10\                       \7\
Alpha emitting nuclides, but no neutron emitters,     2 x 10-\1\       5.4 x  9 x 10-\5\   2.4 x 10-   1 x 10-\1\    2.7 x 10-    1 x 10\3\    2.7 x 10-
 are known to be present \a\........................                   10\0\                     \3\                      \12\                       \8\

[[Page 29016]]

 
Neutron emitting nuclides are known to be present or  1 x 10-\3\   2.7 x 10-  9 x 10-\5\   2.4 x 10-   1 x 10-\1\    2.7 x 10-    1 x 10\3\    2.7 x 10-
 no relevant data are available.....................                     \2\                     \3\                      \12\                       \8\
--------------------------------------------------------------------------------------------------------------------------------------------------------
\a\ If beta or gamma emitting nuclides are known to be present, the A1 value of 0.1 TBq (2.7 Ci) should be used.

* * * * *

    Dated at Rockville, Maryland, this 10th day of May, 2013.

    For the Nuclear Regulatory Commission.
Andrew L. Bates,
Acting Secretary of the Commission.
[FR Doc. 2013-11552 Filed 5-15-13; 8:45 am]
BILLING CODE 7590-01-P