[Federal Register Volume 60, Number 188 (Thursday, September 28, 1995)]
[Rules and Regulations]
[Pages 50248-50289]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-23538]




[[Page 50247]]

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Part II





Nuclear Regulatory Commission





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10 CFR Part 71



Compatibility With the International Atomic Energy Agency (IAEA); Final 
Rule

  Federal Register / Vol. 60, No. 188 / Thursday, September 28, 1995 / 
Rules and Regulations   

[[Page 50248]]


NUCLEAR REGULATORY COMMISSION

10 CFR Part 71

RIN 3150-AC41


Compatibility With the International Atomic Energy Agency (IAEA)

AGENCY: Nuclear Regulatory Commission.

ACTION: Final rule.

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SUMMARY: The Nuclear Regulatory Commission (NRC) is revising the 
regulations governing the transportation of radioactive material. The 
final rule conforms NRC regulations with those of the International 
Atomic Energy Agency, and codifies criteria for packages used to 
transport plutonium by air. This action is necessary to ensure that NRC 
regulations reflect accepted international standards and comply with 
current legislative requirements.

EFFECTIVE DATE: April 1, 1996. Section 71.52 expires April 1, 1999.

ADDRESSES: Single copies of the regulatory analysis for this rule may 
be obtained on request from the contact. Copies of the regulatory 
analysis may be examined and copied, for a fee, in the Commission's 
Public Document Room, at 2120 L Street (Lower Level), NW., Washington, 
DC.

FOR FURTHER INFORMATION CONTACT: John R. Cook, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, telephone: (301) 415-8521.

SUPPLEMENTARY INFORMATION:

Background

    The U.S. Nuclear Regulatory Commission is revising its regulations, 
for the safe transportation of radioactive material to make them 
compatible with those of the International Atomic Energy Agency (IAEA) 
and to incorporate new criteria for packages used to transport 
plutonium by air. The revised rule, in combination with a corresponding 
amendment of Title 49, Code of Federal Regulations, by the U.S. 
Department of Transportation (DOT), would bring U.S. regulations into 
general accord with IAEA regulations (Regulations for the Safe 
Transport of Radioactive Material, 1985 Edition, Safety Series No. 6). 
The final rule also adopts approval criteria for packages used to 
transport plutonium by air. These criteria were developed in response 
to Public Law 94. Except for these revisions, NRC's basic standards for 
packaging and transportation remain essentially unchanged. These 
regulations apply to all NRC licensees who transport, or offer for 
transport, byproduct, source, or special nuclear material, and will 
help ensure the continued safe transportation of radioactive materials 
in domestic and international commerce.
    In addition, three Petitions for Rulemaking, concerning the 
transportation of Low Specific Activity (LSA) radioactive material, are 
denied in this action.
    In 1969, the IAEA, recognizing that its international transport 
regulations should be revised from time to time on the basis of 
scientific and technical advances, as well as accumulated experience, 
invited member states to submit comments and suggested changes to the 
regulations. As a result of this initiative, the IAEA issued revised 
regulations in 1973 (Regulations for the Safe Transport of Radioactive 
Material, 1973 Edition, Safety Series No. 6). The IAEA also decided to 
periodically review its transportation regulations, at intervals of 
about 10 years, to ensure that the regulations are kept current. As a 
result, a review of IAEA regulations was initiated, in 1979, that 
resulted in the publication of revised regulations in 1985 (Regulations 
for the Safe Transport of Radioactive Material, 1985 Edition, Safety 
Series No. 6).
    On August 5, 1983 (48 FR 35600) NRC published, in the Federal 
Register a final revision to 10 CFR Part 71, ``Packaging and 
Transportation of Radioactive Material.'' That revision, in combination 
with a parallel revision of the hazardous materials transportation 
regulations of DOT, brought U.S. domestic transport regulations at the 
Federal level into general accord with the 1973 edition of IAEA 
transport regulations. Some of the revisions that were eventually 
included in the 1985 IAEA regulations were anticipated by NRC and DOT 
when they were finalizing their transportation regulations in 1983. 
These changes were incorporated in Titles 10 and 49 of the Code of 
Federal Regulations at that time.
    On June 8, 1988 (53 FR 21550) NRC published a proposed revision to 
its regulations in 10 CFR Part 71 in the Federal Register for the 
purpose of making U.S. transportation regulations compatible with the 
1985 edition of the IAEA regulations. In a parallel rulemaking, DOT 
published a proposed revision to its radioactive material 
transportation regulations on November 14, 1989 (54 FR 47454). Several 
corrections to the NRC proposed rule were published in the Federal 
Register on June 22, 1988 (53 FR 23484). Interested persons were 
invited to submit written comments and suggestions on the NRC proposal 
and/or the supporting regulatory analysis by October 6, 1988. The 
public comment period was subsequently extended to February 9, 1990. On 
December 8, 1994, the NRC staff provided a briefing on the proposed LSA 
requirements and the other revisions at the 416th meeting of the 
Advisory Committee on Reactor Safeguards (ACRS). This meeting also 
provided industry and the public another opportunity to present their 
views on the revisions. Based on the public comments, consultations 
with DOT, and other considerations, the Commission is adopting the 
proposed rule, with some modifications.

Discussion of Major Changes From Current Requirements

    Most of the revisions presented in the proposed rule are being 
adopted in the final rule. These include additional hypothetical 
accident test criteria for certain types of packages, an increase in 
the number of radionuclides with listed A1 and A2 values, 
changes in the currently listed A1 and A2 values for some 
radionuclides, simplification of fissile material transport classes, 
revised requirements for shipment of LSA materials, and inclusion of 
criteria for packages used to transport plutonium by air. These changes 
are discussed in more detail in the following paragraphs.

Additional Accident Test Requirements

    IAEA deep-water immersion and dynamic crush tests are adopted in 
the final rule. The 200 meter (656 ft) deep-water immersion test has 
been added to the requirements for Type B packages (casks) authorized 
for irradiated fuel content in excess of 37 PBq (10\6\ Ci)(Sec. 71.61 
Special requirement for irradiated nuclear fuel shipments). The purpose 
of the deep immersion test, which can be satisfied through engineering 
evaluation or actual physical test (Sec. 71.41), is to ensure that the 
cask containment system does not collapse, buckle, nor allow inleakage 
of water, if submerged at 200 m (656 ft).
    A dynamic crush test (Sec. 71.73(c)(2) Crush) has also been added 
to Type B package requirements, for certain lightweight packages that 
are minimally vulnerable to damage in the 9 m (30 ft) drop test, but 
which have a high potential for radiation hazard, if package failure 
occurs. IAEA regulations require the crush test in place of the 9 m (30 
ft) drop test, for these packages. NRC is requiring both the crush test 
and drop test, for lightweight packages, to ensure that package 
response to both crush and drop forces is within applicable limits.
    These requirements only apply to package designs certified after 
this final 

[[Page 50249]]
rule becomes effective. Further, this rule does not apply to packages 
fabricated under previous versions of Part 71; however, previously 
fabricated packages are subject to multilateral approval, when used for 
international transport (Sec. 71.13(b)).

Expansion of Radionuclide List and Changes in Radionuclide Limits

    Table A-1, in 10 CFR Part 71, Appendix A, lists the Type A package 
quantity limits (A1 and A2 values) for many radionuclides. 
The final rule increases the number of radionuclides listed, from 284 
to 378. The final rule also adopts the revised A1 and A2 
values contained in the 1985 edition of the IAEA regulations. As a 
result, 144 A1 values previously listed in Table A-1 are being 
increased, and 73 are being decreased, while 129 A2 values are 
being increased, and 95 decreased. In addition, the final rule modifies 
the method used to determine A1 and A2 values for unlisted 
radionuclides.

Simplification of Fissile Material Classes

    The final rule revises the criteria for shipment of fissile 
material. Specifically, the rule eliminates the three fissile class 
designations currently used establishes a single set of criteria for 
all packages of fissile material, uses the transport index as the 
primary control for the number of fissile packages that may be 
transported together, and requires special arrangements for fissile 
packages that do not meet the established criteria.

Inclusion of Criteria for Air Shipment of Plutonium

    The final rule amends Part 71 to include approval criteria for 
packages used to transport plutonium by air (Secs. 71.64, 71.74, and 
71.88). These criteria were developed as a result of Pub. L. 94-79, 
which prohibited NRC from licensing the air shipment of plutonium, in 
any form, until NRC certified to the Congress that a safe container had 
been developed. The NRC subsequently developed and certified package 
criteria to Congress and published the criteria in NUREG-0360, 
Qualification Criteria to Certify a Package for Air Transport of 
Plutonium, dated January 1978. This final rule incorporates these 
criteria. There are no corresponding criteria in IAEA regulations.

Modifications From Proposed Rule

    The final rule differs from the proposed rule in several 
significant respects and are described as follows:
    1. Package limit for Shipment of LSA and Surface-Containment-Object 
(SCO) Material. In its 1985 regulations, the IAEA added a limit of 10 
mSv/hour (1 rem/hour) at 3 meters for the radiation level from the 
unshielded contents of LSA and SCO (Surface Contaminated Object) 
packages not designed to withstand accidents. This radiation level 
limit controls the external radiation exposures to individuals if an 
LSA package is severely damaged in a transportation accident.
    The IAEA limit considers the loss of package shielding during an 
accident but it does not consider the possibility that a package's 
contents might be released and redistributed, causing a reduction in 
self-shielding of the contents. The reduction in self-shielding could 
result in potential accident radiation levels that significantly exceed 
IAEA's 10 mSv/hour (1 rem /hour) at 3 meters limit.
    The IAEA dose rate limit provides a significant added degree of 
protection over the 1973 IAEA regulations (which specify no quantity 
limit for LSA packages). NRC and DOT did not believe, however that the 
IAEA limit provided the same level of safety for all types of LSA 
material, particularly for relatively large quantities of radioactive 
materials contained in dispersible LSA materials (e.g., resins and 
other media used in liquid radioactive waste treatment).
    In lieu of the radiation level limit, DOT and NRC proposed a 
2A1 quantity limit for all LSA packages. Although this proposal 
addressed the accident concern by directly limiting package quantity, 
it was not compatible with the IAEA provisions. Both agencies received 
many comments from industry on the proposed 2A1 quantity limit 
that objected to the impacts on occupational dose and shipping costs. 
Further, after a briefing on the draft final rule on December 8, 1994, 
the Advisory Committee on Reactor Safeguards (ACRS) issued a letter 
report, dated December 19, 1994, recommending, inter alia, that the 
requirements again be reevaluated with the objective of making them 
equivalent to the IAEA regulations.
    After consideration of comments from ACRS and industry, DOT and NRC 
have agreed to adopt the IAEA LSA provisions. Accordingly, the final 
rule imposes a limit on the external radiation level at 3 meters from 
the unshielded contents of LSA-II, LSA-III, or SCO-II packages of 10 
mSv/hour (1 rem/hour) (Sec. 71.10(b)).
    2. The final rule delays imposing the LSA package external 
radiation level limit for 3 years. The effect of imposing the LSA 
package limit is to reduce the quantity of LSA materials that can be 
transported in non-Type B, LSA packages. The final rule may increase 
demand for Type B packages, and there are very few currently available. 
NRC had proposed a 1 year delay in implementing the new LSA rules. 
Industry comments expressed the view that 1 year is not an adequate 
period of time to design a package, have it approved by NRC, and 
manufacture a reasonable number of Type B waste packages. NRC agrees, 
and has included a delay of 3 years from the effective date of this 
rule for implementation of this provision of the final rule 
(Sec. 71.52).
    3. The proposed rule would have adopted 2A1 as the threshold 
below which licensees are exempt from NRC requirements for packages 
containing LSA material (except for Secs. 71.5, 71.88 and 71.53). 
Because NRC and DOT are adopting the IAEA LSA package limit, the final 
rule changes the exemption threshold to 1 rem/h at 3 m 
(Sec. 71.10(b)(2)). Thus, designs for packages used to ship LSA or SCO 
in quantities where the external dose rate exceeds 1 rem/h at 3 m from 
the unshielded material will be subject to NRC Type B package 
regulations. Package designs for lesser quantities of LSA or SCO will 
be self-certified, by package designers, as meeting applicable DOT IP-
1, IP-2, IP-3, Type A, or strong tight, package regulations. [Licensees 
should note that DOT has prescribed, in its final rule, the use of IAEA 
Industrial Packages (IP-1, IP-2, and IP-3) for LSA and SCO material. 
For domestic transportation only, DOT also provides for the use of Type 
A, and strong tight, containers.]
    4. For compatibility with IAEA and DOT requirements, a new, 
``Sec. 71.77 Qualification of LSA-III Material,'' has been added to 
Subpart F. This section prescribes assessment of LSA-III material 
leaching. (In the proposed rule, Sec. 71.77 contained ``Tests for 
special form radioactive material.'' Those requirements have been moved 
to Sec. 71.75 ``Qualification of special form material,'' in the final 
rule.)

Other Administrative Actions

    The final rule corrects numerical errors in Secs. 71.20(b)(3) and 
71.24(b)(4) of the current rule (Secs. (71.20(c)(3) and 71.24(c)(4), 
respectively, of the proposed rule). These errors, which were not 
identified at the time the proposed rule was published, resulted when 
the limit for graphite was expressed as an atomic ratio, instead of a 
mass ratio. The errors were inadvertently adopted, in Part 71, during a 
rulemaking in 1983, to make 

[[Page 50250]]
NRC regulations compatible with 1973 IAEA transportation regulations. 
IAEA has subsequently corrected these errors in the 1985 edition of its 
transportation regulations.
    Section 71.20(b)(3), as currently written, limits the mass of 
graphite to ``* * * 150 times the total mass of uranium-235 plus 
plutonium.'' Section 71.20(c)(3), in the final rule, would be amended 
to read as follows: ``The total mass of graphite present does not 
exceed 7.7 times the total mass of uranium-235 plus plutonium.'' 
Section 71.24(c)(4) would be similarly revised to change the limits on 
graphite from 150 to 7.7 times the total mass of uranium-235 plus 
plutonium.
    NRC is correcting these errors in this final rule. The affected 
sections may bear on the criticality safety of fissile materials in 
transport. In addition, these corrections are expected to have minimal 
impact because there are no shipping casks currently being used that 
were designed using the erroneous provisions.

Summary and Resolution of Public Comments

    There were 171 letters of comment received on the proposed rule 
from industry, State, and local governments; environmental 
organizations; medical facilities; and members of the public. A 
discussion of general comments is presented below, followed by 
responses to comments on specific sections of the proposed rule.
    One of the most frequent comments noted differences among NRC, DOT, 
and IAEA definitions and requirements where there were no reasons for 
the differences. Many of the differences between NRC and DOT 
requirements resulted from the long period of time between publication 
of the NRC proposed rule (June 8, 1988) and publication of the DOT 
proposed rule (November 14, 1989; 54 FR 47454). The two proposed rules 
were intended to be published on or about the same date but 
circumstances did not permit concurrent publication. Between 
publication of the NRC and DOT rules, IAEA published a complete set of 
minor changes and changes of detail to its regulations. These changes 
were not contained in the NRC proposed rule, but were introduced in the 
DOT proposed rule. In addition, a large number of printing errors 
appeared in the text of the NRC proposed rule. Only the most 
significant errors were rectified in a correction notice published June 
22, 1988 (53 FR 23484). The remaining inconsistencies have been 
corrected in the final rule.
    Another frequently raised comment was in response to NRC's 
inclusion of new criteria for the air transportation of plutonium. Out 
of 171 total letters of comment on the proposed rule, 119 of those 
letters were concerned with the single issue of air transportation of 
plutonium. In general, these letters requested that NRC codify the 
NUREG-0360 criteria for the safe air transportation of plutonium, 
notwithstanding urging by the U.S. Department of Energy (DOE) that NRC 
withhold codification until it could consider rules being developed by 
IAEA for the safe air transportation of plutonium. Many of these 
letters, primarily from residents of Alaska, attributed development of 
the NUREG-0360 1 criteria to U.S. Senator Frank Murkowski. 
However, the criteria in NUREG-0360 were developed by the NRC in 
response to Public Law 94-79, enacted in 1975. (Senator Murkowski 
sponsored much more recent legislation on transportation of plutonium 
by air, identified as Section 5062 of Public Law 100-203, for which 
regulatory criteria have not been developed.) NRC has relied on the 
NUREG-0360 criteria for plutonium transportation by air since the 
criteria were published in 1978. DOE's request that NRC withhold the 
codification of the NUREG-0360 criteria while NRC considers the IAEA 
alternative cannot be accommodated because there is no existing IAEA 
alternative to consider and none is expected for several years. 
Although the IAEA development process has begun, the process is long 
and multifaceted. Predictions as to final content of an IAEA 
alternative cannot be made at this time. It also should be noted that, 
under Public Law 94-79, the proposed criteria would apply to any U.S. 
import, export, or domestic plutonium air transport regardless of IAEA 
regulations. Accordingly, the plutonium air transport criteria are 
incorporated in the final rule.

    \1\ Copies of NUREG-0360 may be purchased from the 
Superintendent of Documents, U.S. Government Printing Office, P.O. 
Box 37082, Washington, DC 20013-7082. Copies are also available from 
the National Technical Information Service, 5285 Port Royal Road, 
Springfield, VA 22161. A copy is also available for inspection and 
copying for a fee in the NRC Public Document Room, 2120 L Street, 
NW. (Lower Level), Washington, DC.
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Section 71.0  Purpose and Scope

    One comment suggested that Sec. 71.0 (a) could be clarified by 
referring to the need for a Type B package rather than to licensed 
material in excess of a Type A quantity. Section 71.0 (a)(2) would then 
read ``Procedures and standards for NRC approval of packaging and 
shipping procedures for fissile material and for other licensed 
material required by this Part to be transported in a Type B 
packaging.''
    Although the suggested wording may be a good description of Part 
71, Fissile Type A packages are still subject to NRC approval. 
Therefore a scope based on quantity of radioactive material is better 
than a scope based on a single type of package.

Section 71.4  Definitions

    One comment noted that the term ``licensed material'' is used in 
Part 71, in several locations, but is not defined in Part 71. In 
response to this comment, NRC has added the definition of ``licensed 
material,'' as codified in 10 CFR Part 39, to the definitions in Part 
71. The term ``licensed material'' only includes radioactive material 
licensed by the NRC. One comment noted that in defining the term 
``exclusive use,'' the parenthetical note ``* * * also referred to in 
other regulations as `sole use' or `full load' '' is no longer 
necessary. Those other terms have been almost completely phased out, 
and IAEA has eliminated the clarifying note. NRC agrees and also has 
eliminated the clarifying note.
    One comment noted that the definition of ``exclusive use'' requires 
that loading and unloading be performed by personnel having 
radiological training and resources appropriate for safe handling of 
the consignment. However, the definition provides no criteria to 
indicate what that training should be. NRC believes this is an area 
where the regulation includes a sufficient level of detail to define 
the intent of the provision. NRC further notes that DOT has established 
requirements for hazardous material employee training (see 49 CFR Part 
172, Subpart H, Secs. 172.700-172.704, effective July 2, 1992).
    One comment suggested that the term ``transport index'' specify 
that the number be rounded up ``to the next tenth'' rather than ``to 
the first decimal place.'' NRC believes that either terminology is 
adequately clear, and is retaining the original wording for uniformity. 
This wording has been used satisfactorily over a number of years.
    One comment suggested that the ``Natural uranium'' definition 
should be clarified to indicate that the phrase ``the remainder being 
uranium-238'' refers strictly to a weight basis, not to a radioactivity 
basis. NRC has made the clarification.
    One comment raised the question whether ``licensee'' and ``licensee 
of the Commission'' are synonymous, and whether the terms include 
``persons 

[[Page 50251]]
licensed by an Agreement State,'' so that the general licenses of 
Secs. 71.12-71.24 could apply. NRC asserts that the terms ``licensee'' 
and ``licensee of the Commission'' are synonymous. For uniformity, the 
NRC has eliminated the longer of the two terms in the final rule. 
Neither term includes Agreement State licensees. However, Agreement 
State licensees engaging in activities in non-Agreement States, or in 
offshore waters, under the reciprocity provisions of 10 CFR Part 150, 
``Exemptions and Continued Regulatory Authority in Agreement States and 
in Offshore Waters under Section 274,'' are subject to the requirements 
of 10 CFR Part 71. In such instances, the NRC general licenses 
mentioned above apply to Agreement State licensees.
    One comment noted that the term ``specific activity'' should only 
be used when describing the radioactivity of a radionuclide per unit 
mass of the element. When describing the radioactivity per unit mass of 
a material in general, the comment suggested the use of the words 
``concentration of radioactivity.'' NRC has been unable to confirm any 
preferred limited use of the term ``specific activity,'' and, in view 
of the years of successful international use of the term in its broader 
sense, plans to continue that broader use.
    One comment noted that the NRC and DOT definitions of ``exclusive 
use'' are not identical, and that the DOT definition appears 
preferable. In the final rules promulgated by NRC and DOT, the 
definitions of ``exclusive use'' are identical.
    One comment noted a difference in quantities, for DOT's proposed 
rule ``highway route controlled quantities,'' in 49 CFR 173.403, and 
for NRC's ``advanced notification of shipment of nuclear waste'' 
requirements in 10 CFR 71.97. The limits were intended to be the same. 
As the comment suggested, the error (by NRC) was caused by the rounding 
of the International System (of units) (SI) and customary units and has 
been corrected in this final rule.

Section 71.4  Definitions (Dual Unit System--The International System 
of Units Followed or Preceded by U.S. Standard or Customary Units).

    Ten comments suggested both support for the dual unit system used 
in both NRC and DOT proposed regulations and potential problems that 
might result from a dual unit system. Several other comments suggested 
that NRC and DOT be consistent in the use of units. NRC and DOT intend 
to use dual units in specifying the regulatory requirements. The 
introductory language to Sec. 71.4 states that the different units are 
functionally equivalent and can be used interchangeably for purposes of 
this part. There are no paperwork requirements in Part 71 (e.g., 
records, reports) where the mandatory use of units is specified. DOT 
regulations also specify regulatory requirements in terms of dual 
units. In 49 CFR 171.10, DOT specifies that the SI units are intended 
to serve as the standard, but that the customary units (rounded) are 
included to provide a functionally equivalent limit. The dual unit 
approaches used by NRC and DOT are compatible.
    In addition, DOT specifies, in 49 CFR Part 172, the units that must 
be used to satisfy the communication standards for shipping papers and 
package labels. Sections 172.203(d)(4)and 172.403(g)(2) require that 
shipping papers and package labels be completed either in SI units 
alone or in SI units and customary units. These requirements also 
permit, for a period of one year after the effective date of the final 
rule, the use of customary units on shipping papers and package labels 
for domestic shipments only.
    One comment noted that the double conversion from customary units 
to SI units, and back to customary units produces specifications that 
are out of line with standard material sizes. For example, a test with 
what was a standard 6-inch-diameter mild steel bar, with an edge radius 
of \1/4\ inch, was proposed as a test with a 5.91-inch diameter mild 
steel bar, with an edge radius of 0.236 inch. The converted customary 
units of length and weight have been returned to their original values 
in the final rule.
    One comment suggested greater consistency of units between the NRC 
and DOT transportation regulations and the Commission's ``Standards for 
Protection against Radiation'' in 10 CFR Part 20. Since the NRC and DOT 
transportation rules were proposed, NRC has revised 10 CFR 20.1004, 
``Units of Radiation Dose,'' and 10 CFR 20.1005, ``Units of 
Radioactivity,'' to permit the use of either customary or SI units, 
These revisions achieve greater consistency of units among 
transportation and radiation protection regulations.
    One comment noted that differences between IAEA and Part 71 A 
values (expressed in conventional units) may cause problems in 
international transport. The curie values in Safety Series #6, Table I 
are approximate, rounded down from the TBq values after conversion to 
Ci, whereas the curie values in Table A-1 Part 71 are converted from 
the TBq values to three significant figures without rounding down. The 
Part 71 method was used because it yields values that more closely 
approximate previous Table A-1 values. As noted earlier in this 
preamble, DOT regulations will require the use of the SI units in 
shipping papers and labels for international shipments (although 
conventional units may be used in addition to the SI units). The use of 
SI units should retain consistency with the IAEA regulations.
    One comment suggested that the term ``transport index'' be defined 
using both customary and SI units, as IAEA has done. The proposed 
definition was expressed only in customary units. NRC agrees with this 
suggestion and has adopted the DOT definition of ``transport index'' 
which includes both customary and SI units.

Section 71.4  Definitions (LSA and SCO in Particular)

    Several comments related to clarification of LSA definitions.
    Two comments noted the typographical error in the proposed rule in 
which the ``water with tritium'' concentrations for LSA-II were printed 
as 27.0 Ci/ (1 TBq/), rather than as 27.0 Ci/l (1 
TBq/l). Two other comments noted that the numerical values differed 
from those in the DOT proposed rule (20 Ci/l and 0.8 TBq/l, 
respectively). One comment stated a preference for the 27.0 Ci/l limit.
    NRC values in the proposed rule were derived from the IAEA and DOT 
values by rounding up the terabequerel limit and then converting to 
curies. For consistency, NRC has adopted the IAEA and DOT values in the 
final rule.
    Three comments were concerned with the definition of LSA-I. The 
first comment noted that material generated from the extraction of 
uranium or thorium was not classified into any LSA category. The 
comment recommended an LSA-I classification for this material. Another 
comment recommended that the term ``contaminated earth'' in LSA-I be 
expanded to include ``soil, earth, concrete rubble, and other bulk 
debris.'' A third comment expressed concern that mill tailings 
exceeding 10-6 A2/g could not be shipped in bulk under the 
proposed rule. The comment recommended that either mill tailings be 
specifically included in the definition of LSA-I without an activity or 
concentration limit, or the specific activity limit for LSA-I be 
increased to 4x10-6 A2/g.
    NRC agrees that ore-like materials (materials with highly uniform 
distribution of small quantities of radionuclides) should be 
transported as LSA-I material. Accordingly, the definition of LSA-I has 
been changed from ``contaminated earth * * * `` to 

[[Page 50252]]
``contaminated earth, mill tailings, concrete rubble and other bulk 
debris * * *'' Further, NRC believes that mill tailings will meet the 
proposed 10-6 A2/g specific activity limit, and therefore has 
not increased the limit.
    Two comments suggested that NRC include a definition of the term 
``closed transport vehicle'' used in the definition of LSA-I. This term 
has been removed from the definition of LSA-I because NRC and DOT 
concluded the use of a vehicle-based term in the definition of a 
material was inappropriate. ``Closed transport vehicle'' is defined in 
DOT's rule (49 CFR 173.403(c)).
    One comment suggested that LSA-II material definition be expanded 
to include activated materials, consolidated wastes, and materials 
intrinsically contained in a relatively insoluble matrix. LSA-II is 
expected to include primarily unsolidified material in which the 
radioactive material may or may not be uniformly distributed, including 
lesser activity resins and filter sludges, other similar materials from 
reactor operations, similar materials from other fuel cycle operations, 
scintillation vials, and hospital, biological, and decommissioning 
wastes. There is, however, no prohibition against activated materials, 
consolidated wastes, and materials intrinsically contained in a 
relatively insoluble matrix in group LSA-II, provided the specific 
activity limit is met. The IAEA established the LSA-III group 
principally for irradiated reactor parts and other activated, or 
activated and contaminated, equipment that exceed the limits for the 
other LSA groups. NRC does not believe it is necessary to expand the 
LSA-II group definition to include these materials. The NRC believes 
that to do so might cause confusion with the LSA-III definition.
    One comment stated that dewatered material should be defined as a 
solid for LSA-II. NRC agrees that dewatered resins should be subject to 
the specific activity for solids under LSA-II and notes that there is 
no prohibition against dewatered resins in LSA-II.
    One comment asked whether the specific activity limits for LSA-II 
and LSA-III materials were pre- or post-solidification. The specific 
activity limits apply to materials as prepared for shipment, i.e., 
post-solidification. However, licensees should note that packaging or 
shielding material may not be considered in determining either the 
specific activity or the radiation level at 3 m.
    One comment recommended that NRC remove the criterion for leaching 
that is applicable to LSA-III solids. The criterion limits the loss of 
radioactive material per package, when the package is placed in water 
for 7 days, to 0.1 A2. Another comment stated that the criterion 
for leaching in the definition of LSA-III needed to be compatible with 
the leachability index requirements for solidified waste in 10 CFR 
Parts 60 and 61.
    A control on the potential intake of these LSA-III materials is 
necessary because the radioactivity is not entirely insoluble. Because 
non-Type A packaging might be used in transporting these materials, a 
release of 10-2A in an accident is assumed, with a possible 
bystander uptake of 10-3 A2, under the standard model for 
determining A2 values. Because the total body uptake must be 
limited to 10-6 A2, the package's dispersible radioactive 
contents (i.e., the leachate liquid), must not exceed 0.1 A2. For 
purposes of compatibility with IAEA and DOT requirements, a new 
Sec. 71.77, ``Qualification of LSA-III Material,'' has been added to 
Subpart F. This section prescribes testing requirements for assessment 
of LSA-III material leaching. The hazard from the transportation of 
these materials is different from that posed by their disposal; 
therefore, no attempt has been made to achieve compatibility between 
transportation and disposal leachability limits.
    One comment found the proposed rule unclear on the need for three 
LSA categories and how to classify materials under the criteria, 
including compacted dry active waste. IAEA developed the three LSA 
groups to differentiate controls based on the activity, distribution, 
and form of LSA material. The LSA-I group accommodates very uniformly 
distributed materials, such as ores. LSA-III accommodates large 
activated parts or solidified materials. LSA-II accommodates less 
uniformly distributed materials, such as compacted dry active waste.
    One comment described radioactive atoms in activated products as 
inherently non-dispersible and relatively non-leachable. The comment 
recommended that activated materials be authorized for shipment as LSA-
I, provided other transportation requirements are met. Although 
activated materials do not pose a dispersibility hazard, these 
materials are subject to localized concentrations of non-uniformly 
distributed material. Consequently activated materials are included in 
groups LSA-III and LSA-II.
    One comment suggested changing the definition of SCO from ``* * * 
not itself radioactive * * *'' to ``* * * not classed as radioactive 
material under these rules * * *,'' since nothing is free of 
radioactive material. NRC and DOT have adopted this comment.
    Several comments identified a typographical error in the limit for 
non-fixed contamination from beta and gamma emitters on the accessible 
surface of SCO-I objects. That value has been changed from 1.08  x  
10-5 Ci/cm \2\ to 10-4 microcurie/cm \2\. These comments also 
noted inconsistencies in the NRC and DOT contamination limits e.g., 
(1.08  x  10-4 Ci/cm \2\ and 10-4 microcurie/cm \2\, 
respectively). NRC has adopted the DOT convention for these limits in 
the final rule.
    One comment inquired as to whether it was consistent for NRC not to 
exempt SCO-I from transportation requirements when facilities with 
similar contamination levels may be released for unrestricted use 
according to NRC Regulatory Guide 1.86. Under the final rule, SCO-I 
group materials are exempt from NRC regulations, except for one 
Sec. 71.5 requirement that licensees comply with DOT requirements. 
Further, the SCO-I non-fixed surface contamination limits are greater 
than, not similar to, the corresponding acceptable surface 
contamination levels in Table 1 of NRC Regulatory Guide 1.86.
    Several comments noted that the term ``inaccessible surface'' used 
in the SCO-I definition is not defined and that it was not clear how to 
comply with a limit for surfaces that were inaccessible. This provision 
provides for the disposal of materials that have contaminated surfaces 
that are not readily accessible. Examples of inaccessible surfaces 
include: inner surfaces of pipes, inner surfaces of maintenance 
equipment for nuclear facilities, and inner surfaces of glove boxes. 
Compliance can be achieved by sampling a small area of the surface that 
may be accessible or by a documented estimate of the inaccessible 
surface contamination.
    One comment stated a belief that the implementation of SCO groups 
would: (a) Further complicate the preparation and shipment process, 
without an increase in the safety and quality of waste shipments; (b) 
result in a significant increase in personnel exposure costs, and 
delays for preparation and disposal of radioactive waste; (c) require 
substantial initial personnel training; and (d) require extensive 
revisions of existing procedures and waste shipping computer programs. 
NRC acknowledges that the introduction of multiple LSA and SCO groups 
complicates the transportation of LSA materials. The IAEA consensus was 
that it was appropriate to regulate SCO separately from LSA materials. 
The purpose of 

[[Page 50253]]
these groups is to recognize the lesser hazard of LSA and SCO relative 
to other radioactive materials, and to provide relief from shipment 
requirements that would otherwise apply to these materials, while still 
assuring safety.
    With regard to exposure, it is true that the LSA groups will 
require some increased material treatment or handling. However, this 
handling is necessary to eliminate the current practice in which there 
is no quantity limit on LSA packages. This situation poses a risk to 
the public during transport. Costs will increase, but not by an amount 
considered significant for the industry. Training with regard to the 
LSA groups, or any new provision, will be required. Periodic training 
of hazardous material employees regarding the safe transportation of 
hazardous materials is required by DOT regulations (49 CFR Part 172 
Subpart H); instruction with regard to the LSA and SCO groups may be 
included at that time.
    Implementing the LSA groups will require revision of procedures and 
computer codes. These costs are judged to be acceptable in order to 
achieve compatibility with the IAEA regulations for the safe transport 
of radioactive materials.
    A comment noted that the SCO classification ``appears to be well-
meaning,'' but that the proposed criteria (presumably the proposed 
2A1 limit) ``detract from its potential benefit and utility,'' and 
that it would be easier and less expensive for both producers and 
consumers of electricity to enjoy the benefits of new transportation 
systems without the related restrictions. As stated previously, NRC has 
adopted the IAEA 10 mSv/h (1 rem/h) at 3 m limit for LSA packages, and 
believes that a limit is needed to protect the public from the 
potential for excessive external radiation exposure in the case of a 
severe transportation accident.
    One comment suggested that the rule make clear that not every SCO 
needs to be surveyed and that a random representative survey is 
adequate. There is no requirement that each SCO in a package be 
surveyed. The shipper must be able to demonstrate, however, that the 
package contents comply with applicable SCO definitions.
    One comment objected to the upper limit for removable surface 
contamination for SCO-II (10-2 Ci/cm \2\ for beta and 
gamma emitters) because this limit is a factor of 90 less than current 
LSA limits, and would require extensive decontamination of reactor 
outage equipment at each site. The comment stated such decontamination 
is not warranted because it violates the as low as reasonably 
achievable (ALARA) principle, and is not justified based on shipping 
experience. The comment suggested that an SCO-III group be defined for 
materials exceeding SCO-II, and that Type A packaging be required for 
such materials.
    Apparently, this comment is comparing the SCO-II limit for 
removable (non-fixed) surface contamination with the current LSA limit 
that applies to nonradioactive material objects that are externally 
contaminated with radioactive material that is not readily dispersible. 
The SCO-II limit for fixed surface contamination is a more appropriate 
comparison with the current limit for not readily dispersible 
contamination. The SCO-II fixed contamination limit is 20 times greater 
than the current LSA limit for not readily dispersible contamination.

Section 71.5  Transportation of Licensed Material

    Two comments asked for clarification of the specification ``* * 
*outside of the confines of its plant or other place of use,'' when 
describing transportation made subject to DOT regulations. One of those 
comments suggested that the provision be reworded as ``* * *outside the 
site of usage, as specified in the NRC license, or where transport is 
on public highways.'' This wording clarifies the provision and has been 
included in the final rule. Similar wording has been substituted in 
Sec. 71.0(c).
    A comment asked whether Sec. 71.5(b) means ``that an approval must 
be obtained when the shipment is covered by local State regulations and 
those regulations will be followed.'' The purpose of Sec. 71.5(b) is to 
impose, by NRC authority, pertinent DOT requirements on shipments, by 
NRC licensees, that are not normally subject to DOT requirements. There 
is no exemption from the requirement of Sec. 71.5(b) regarding 
compliance with State or local regulations.

Section 71.10   Exemption for Low Level Materials

    A comment noted that the SI unit specification of 74 kBq/kg 
(0.002Ci/g) for exempted low-level radioactive material in 
Sec. 71.10(a) is not consistent with the 70 Bq value specified in the 
DOT proposed rule. The specification in Sec. 71.10(a) has been changed 
to 70 Bq/g, the value in the DOT's final rule. This exemption is 
applicable only with respect to transportation, and is not generally 
applicable to other Commission-regulated activities.
    A comment noted that it would be useful to have an exemption for 
small quantities of radioactive material in Sec. 71.10(a) as well as 
the exemption for LSA material. The safety rationale developed by IAEA 
2 for LSA material does not extend to other radioactive materials. 
IAEA has been informed that a small quantity exemption may be a useful 
concept. However, this exemption has not been developed yet.

    \2\ International Atomic Energy Agency Safety Series #7--
``Explanatory Material for the IAEA Regulations for the Safe 
Transport of Radioactive Material'' (1985 Edition). Available from 
Bernam-Unipub, 4611-F Assembly Drive, Lanham, MD 20706-4391. Tel. 
(301) 459-7666.
---------------------------------------------------------------------------

    One comment asked that NRC clarify the use of a reference to 
Sec. 71.53 in the ``Exemption for low-level materials'' provision of 
Sec. 71.10(b), a provision that pertains to Type A and LSA packages. In 
addition to control over excessive radiation, the Commission's 
responsibility with respect to fissile material is to provide 
reasonable controls to avoid the occurrence of accidental criticality. 
The regulatory standards for this are found in Secs. 71.55 and 71.59. 
There are some relatively common types of fissile material packages for 
which there is no credible risk of criticality in transport, even in 
the absence of controls. These packages are described in Sec. 71.53, 
and are exempted from the criticality controls of Secs. 71.55 and 
71.59, because the controls are unnecessary.
    The provisions of Sec. 71.10, ``Exemption for low-level 
materials,'' provide broad exemptions from 10 CFR Part 71 rules that 
relinquish to DOT the control of types of shipments that are of low 
risk both from radiation and criticality standpoints. To ensure that 
only low criticality risk shipments are included in Sec. 71.10(b), NRC 
restricts the exemption to Type A and LSA packages that either contain 
no fissile material or satisfy the fissile material exemptions in 
Sec. 71.53. It should be noted that the exemption does not relieve 
licensees from DOT transportation requirements by reason of NRC 
authority, nor does the exemption relieve licensees from the 
restrictions on air transportation of plutonium imposed by Congress.
    The proposed rule introduced a 2A1 quantity limit, for LSA 
packages not designed to withstand accidents (non-Type B packages), to 
control potential external radiation exposures. Thirty comments were 
received requesting that the limit be changed in the final rule. Two 
comments supported no limit; nine supported the IAEA dose limit of 10 
mSv/h (1 rem/h)r at a distance of 3 meters for an unshielded package; 4 
supported higher multiples of A1; and 15 supported the optional 
use of either the IAEA limit or a higher multiple of A1. As 
described previously in this 

[[Page 50254]]
preamble, NRC and DOT have decided that the best overall response on 
the LSA issue and these comments is to drop the proposed 2A1 
quantity limit, and to adopt the IAEA radiation level limit of 10 mSv/h 
(1 rem/h) at 3 m from the unshielded contents.
    One comment suggested that the need for labels on LSA packages 
should be reconsidered. Package labelling falls under DOT jurisdiction. 
In its final rule, DOT has retained the exception from package marking 
and labeling requirements for domestic LSA shipments consigned as 
exclusive use (see 49 CFR 173.427).
    One comment expressed concern over the transition of control of 
packages for shipping Type B quantities of LSA radioactive material 
from NRC to DOT. NRC has a centralized package design approval 
authority, whereas DOT authority allows a shipper to determine 
acceptable package designs (i.e., self-certify package designs). The 
comment expressed apprehension about permitting each shipper to review 
package and shipping restrictions against DOT regulations, a situation 
that could result in some confusion and different interpretations of 
the regulations.
    In the final rule, the IAEA limit of 1 rem/h at 3 m from the 
unshielded material contents has been established as the threshold for 
NRC regulation of LSA or SCO package designs. NRC will review and 
approve, if adequate, designs for packages that contain quantities of 
LSA or SCO material that exceed that limit. The review by regulatory 
authority of package designs for quantities that exceed the IAEA limit 
is consistent with the approach used by other IAEA member states.

Section 71.13  Previously Approved Package

    One comment proposed that the date specified in Sec. 71.13(b)(2) be 
December 31, 1990, instead of December 31, 1992, to be consistent with 
IAEA transportation regulations. The original 1985 IAEA transport 
regulations specified December 31, 1990, as the cutoff date for the 
routine use of packages manufactured under the 1973 edition of the 
regulations. That date was subsequently extended for 2 years by one of 
the periodic updates of IAEA regulations and was properly used in the 
proposed rule. However, since the proposed date of December 31, 1992, 
has passed, the final rule has been revised (by eliminating reference 
to any particular date) to make this provision effective on the date 
that the final rule becomes effective.
    Two comments noted that the preamble to the proposed Part 71 
indicated that Type B and fissile packages fabricated before a certain 
date and not used internationally could continue to be used 
domestically until the end of their useful lives. The licensee would 
not need to demonstrate that the packages satisfy the new crush test or 
deep-immersion test. The comments would take that provision one step 
further and require the crush and deep-immersion tests only for 
international use packages.
    NRC believes that the international package standards should be 
used by the United States for both domestic and international 
shipments, to the extent practicable. However, based on a history of 
safe use under earlier safety standards, and the absence of unfavorable 
operational data, NRC will allow the continued use of existing packages 
in domestic transport until the end of their useful lives. NRC will not 
allow, however, the continued fabrication of packages to the old 
designs. This action permits use of existing packages. It does not 
perpetuate package designs that can be discarded or upgraded to satisfy 
the new standards.
    Another comment suggested grandfathering the existing Type A casks 
now approved for transporting Type B quantities of LSA radioactive 
material, until the Type B waste casks required to satisfy the new 
standards become available. NRC has adopted the suggestion, extending 
the proposed provisions in Sec. 71.52, ``Exemption for low-specific-
activity (LSA) packages,'' to a 3 year period, to give the industry 
time to design, receive approval, and fabricate new Type B waste 
packages.

Section 71.22   General license: Fissile Material, Limited Quantity, 
Controlled Shipment

    One comment requested clarification as to whether the Type A limit 
imposed in Sec. 71.22(c) also applies to Sec. 71.22(d).
    The requirements of Secs. 71.22(a) through 71.22(e) are cumulative, 
each imposing additional requirements on the use of the general 
license. The radioactivity limit and mass limits of Sec. 71.22(c) apply 
to packages, whereas the mass and mass ratio limits of Sec. 71.22(d) 
apply to shipments.
    A comment noted an error, in Sec. 71.22(d)(3), which changed the 
intent of the section. The commenter suggests that the phrase ``exceeds 
unity'' at the end of Sec. 71.22(d)(3) be replaced by the phrase ``does 
not exceed unity.'' NRC agrees and has made that change.

Section 71.24  General License: Fissile Material, Limited Moderator, 
Controlled Shipment

    One commenter asked if the statement in Sec. 71.24(b), ``* * * a 
quality assurance program approved by the Commission as satisfying the 
provisions of Subpart H of this part,'' is any different from ``* * * a 
quality assurance program approved by the Commission.'' The two 
statements are different in that the first is more specific and 
provides more detail. There are several different quality assurance 
programs, in different licensing areas, approved by the Commission. 
Specifying that the program must satisfy Subpart H makes it clear as to 
the type of quality assurance program is required.
    One commenter recommended inserting ``by weight'' after ``1 
percent'' in Sec. 71.24(c)(6). NRC agrees and has made this change in 
Sec. 71.24(c)(7), as well.
    With respect to a general license for a package containing fissile 
contents, one commenter requested clarification of what is meant by 
``no uranium-233'' in Sec. 71.24(c)(6). For a general license under 
Sec. 71.24(c)(6), a package containing fissile contents must have no 
detectable U-233. The method for making this determination can be 
decided by the licensee. For example, the licensee can make this 
determination by performing an assay or by knowing the history of the 
material.

Subpart D--Application for Package Approval

    One comment suggested changing the title of Subpart D to 
``Application for Type B Package Approval'' for clarity. Because NRC 
also approves Type A packages for fissile material, the title of 
Subpart D continues to refer to ``Package Approval.''

Section 71.38  Renewal

    One comment suggested that NRC provide some administrative 
acknowledgment when a timely application for renewal of a certificate 
of compliance has been received to provide proof that timely renewal is 
in effect. The Commission does not believe that proof of timely renewal 
is particularly important and that providing an acknowledgment to each 
registered user of a package would be too burdensome for the benefit 
gained.

Section 71.43   General Standards for All Packages

    Four comments suggested the addition of IAEA regulations relating 
to packaging of liquids and gases to Part 71, including those 
pertaining to the special free drop and penetration tests 

[[Page 50255]]
for liquids and gases. The NRC approves only Type B and fissile 
material packages. The NRC also notes that fissile material packages 
must be evaluated for hypothetical accident conditions more severe than 
the tests for liquids. Furthermore, there are currently no NRC-licensed 
packages designed for gaseous fissile materials and NRC does not 
anticipate any future applications for such packages. These additional 
provisions would complicate regulations that are presently adequate. 
IAEA standards on absorbent material and double containment have been 
selectively included in DOT regulations.
    Eight comments disagreed with the NRC view that Sec. 71.43(f) 
should continue to restrict to ``no significant increase'' any change 
in external surface radiation levels, as a result of subjecting a 
package to the defined normal conditions of transport. The comments 
argued that the 20 percent increase specified in IAEA regulations is a 
safe, reasonable, and practical number that could not reasonably be 
lower, and that specifying a value in the rule provides the package 
design engineer and the NRC review engineer a measurable goal that is 
consistent both with IAEA and with engineering practice.
    Type B and fissile material packages can be readily designed so 
that normal conditions of transport result in no significant increase 
in dose rates, and that a twenty percent increase in dose rates because 
of normal handling is excessive. In addition, if a package were 
designed so that the external dose rate could increase 20 percent 
during normal handling, the package could exceed the dose rate limits 
in Sec. 71.47 during transport, and would be an item of non-compliance. 
NRC and DOT have therefore decided to not adopt the IAEA ``20 percent 
increase'' provision, and to retain the current ``no significant 
increase'' provision.
    Four comments suggest the addition of the special provisions of 
IAEA regulations pertaining to the transportation of radioactive 
material by the air mode. NRC has determined that special requirements 
for transport of packages by air should be excluded from Part 71 
because these provisions are properly incorporated in the carrier 
restrictions imposed by the Department of Transportation.
    Two comments suggested that the phrase ``Account must be taken of 
the behavior of materials under irradiation'' be clarified and 
quantified, perhaps in a regulatory guide, or deleted from Part 71. 
Although there is no regulatory guidance now available relating this 
requirement to transportation packages, it is clear that any effects of 
irradiation on materials used in the package must be taken into 
account. These effects could be the accelerated aging or embrittlement 
of elastomers or elastics and may result in requiring a frequent change 
of gaskets, for example.
    One comment suggested the performance requirement of Sec. 71.43(f) 
be changed to include a numerical sensitivity for the requirement that 
there be ``no loss or dispersal of radioactive contents'' as a result 
of subjecting a package to the specified normal conditions of 
transport. The equivalent paragraph in the IAEA regulations for Type A 
packages is paragraph 537, and does not contain a numerical 
sensitivity. Paragraph 548, of IAEA Safety Series #6, is the equivalent 
of 10 CFR 71.51, for Type B package leaktight sensitivity. Both those 
provisions require Type B packages to be leaktight to a sensitivity of 
10-6 A2/h.
    Three comments noted that IAEA no longer prohibits continuous 
venting of packages in its 1985 edition and urged the NRC to allow the 
practice domestically for Type B packages. The commenters argued that 
although NRC took a strong position, in the preamble to the proposed 
rule, that continuous package venting is ``poor engineering practice,'' 
NRC did not explain why. The commenters noted that DOT regulations do 
not prohibit continuous venting for Type A packages, leaving the 
acceptability of continuous venting to be decided by performance 
requirements. The commenters stated that in some cases it would make 
good sense to allow continuous venting to provide pressure equalization 
and discharge of organically generated hydrogen gas.
    NRC is continuing its ban on continuous venting of Type B packages 
for the following reasons:
    1. Venting of a package containment system during normal conditions 
of transport defeats the purpose of the containment system;
    2. It is practical to design packages that do not rely on venting, 
to relieve pressure under normal conditions of transport;
    3. The use of a vent does not necessarily prevent the generation of 
potentially flammable or explosive gas mixtures; and
    4. The reliability of filters under temperature extremes, varied 
operating conditions, and sustained service has not been established.
    Two comments stated that Mo-99/Tc-99m radiopharmaceutical 
generators are open to the atmosphere to allow changes in ambient 
pressure and that the generators do not vent radioactive material. The 
comments recommended that the prohibition against venting be limited to 
venting radioactive material only and that NRC continue current 
practices.
    NRC believes these comments arise from concern over the reduction 
in the A2 quantity for Mo-99 from 20 curies to 13.5 curies in the 
proposed rule. NRC recognizes that the shipment of Mo-99/Tc-99m 
generators is a special case, and is retaining the 20 curie A2 
value for Mo-99, to permit the continuation of current practices.

Section 71.47  External Radiation Standards for All Packages

    NRC used the term ``accessible external surface'' in its proposed 
rule for determining radiation levels on package surfaces, whereas DOT 
used the term ``external surface'' in its proposed rule. Four comments 
argued that the NRC and DOT regulations for radiation level limits on 
package surfaces should be identical. Most believed that a limit on 
accessible surfaces was the more reasonable standard.
    DOT has indicated that it is considering a petition for rulemaking 
to add the word ``accessible'' to its radiation level regulations and 
will consider that complex issue in a separate action. Pending 
completion of the DOT separate action, NRC has deleted the word 
``accessible'' from this section of the final rule but does not intend 
to alter its practices regarding this provision.
    One comment stated that this paragraph tends to be confusing in 
that it establishes a limit of 2 mSv/h (200 mrem/h) for package surface 
radiation levels, yet Sec. 71.47(b)(2) seems to state that packages 
transported on a flatbed trailer can exceed 2 mSv/h (200 mrem/h), 
provided the radiation level at the planar edges of the trailer is less 
than or equal to 2 mSv/h (200 mrem/h).
    Section 71.47 establishes a generally applicable 2 mSv/h (200 mrem/
h) Package surface radiation-level limit. The section further 
establishes that, if a package is shipped as exclusive use, the 
radiation level may exceed 2 mSv/h (200 mrem/h), provided the 
applicable provisions of paragraphs (a) (with repect to Transport 
Index) through (d) are met. Paragraph (b)(2) restricts the radiation 
level at any point on the vertical planes projected by the outer edges 
of a flat-bed style vehicle to 2 mSv/h (200 mrem/h) (the same limit 
imposed in paragraph (a) for the outer surfaces of closed transport 
vehicles). Thus, provided packages are shipped as exclusive use, 
external radiation levels may exceed 2 mSv/h (200 mrem/h) at the 
surface of packages on flatbed trailers, but not at the outer-edge 
planes of the vehicle. 

[[Page 50256]]


Section 71.51  Additional Requirements for Type B Packages

    One comment suggested that the clarifying provision following 
paragraphs 548(a) and (b) of IAEA regulations be added to Part 71 for 
consistency. The clarifying provision pertains to allowable releases of 
radioactive material from a package containing a mixture of 
radionuclides. This is the case, for example, with spent nuclear fuel 
casks. That clarifying provision has been added.

Section 71.52  Exemption for LSA Packages

    Twelve comments expressed concern that the proposed Part 71 affords 
only a 1-year delay in applying the new LSA rules. NRC established the 
1-year delay to give the industry an opportunity to design and build 
the Type B waste casks that would be required under the new rules. The 
comments uniformly argued that 1 year was not a sufficient period of 
time to design a waste cask, to have it reviewed and approved by NRC, 
and to fabricate an adequate number of casks, to approved designs, that 
satisfy the needs of the new LSA rule. The commenters differed in how 
long they thought that process would take, varying over 2, 3, and 5 
year periods. NRC agrees with the thrust of this comment and has 
established the exemption period at 3 years. Thus existing packagings 
may be used for 3 years and new packagings may be fabricated from 
existing designs for 3 years.
    A consequence of establishing the IAEA LSA/SCO package limit as the 
delineator between NRC and DOT regulation of LSA and SCO packaging [see 
Sec. 71.10(b)(2)] is that, after the 3 year exemption period, LSA will 
be shipped either in DOT authorized packagings, or in NRC certified 
Type B packagings. Accordingly, NRC is discontinuing the practice of 
certifying Type A LSA packages. NRC has therefore not adopted a 
proposed exemption (Sec. 71.52(a)) that only would have applied to NRC 
certification of new Type A LSA package designs.
    One comment stated that the demand for waste casks would rise until 
1993 and then fall again because few of the low-level radioactive waste 
disposal site compacts will permit disposal access. Vendors will 
hesitate to invest in casks that will not be used after 1993 and waste 
will need to be stored onsite.
    NRC is unwilling to accept this proposition and believes that as 
long as NRC specifies the requirements for transportation of waste, 
given adequate time, industry will continue to develop disposal 
options.
    One comment argues that the specific reference to Sec. 71.43(f) 
should be deleted because it is included in the broader reference to 
Secs. 71.41-71.47.
    Section 71.52 exempts exclusive use LSA and SCO packages from the 
additional requirements for Type B packages for a period of 3 years 
from the effective date of the final rule. These LSA packages are still 
subject to other requirements that apply to all packages. The referral 
to these other package requirements includes Secs. 71.41-71.47, plus a 
specific reference to. An argument could also be made for deleting the 
entire reference because those requirements apply regardless of the 
reference in this section. However, NRC chose to include the reference 
in Sec. 71.52 as a reminder that the exemption is only from Sec. 71.51, 
not from all packaging requirements. NRC believes the reference to 
Sec. 71.43(f) (normal conditions of transport tests) is important and 
has decided that it will be retained.
    One comment suggested that SCO be included within the scope of 
Sec. 71.52, and that the 2A1 limit be included in the section for 
clarity. NRC agrees with the comment and has made the clarifications, 
substituting the IAEA LSA limit for 2A1.

Section 71.53  Fissile Material Exemptions

    One comment suggested spelling out the word ``liter'' instead of 
using ``l'' as the abbreviation. Considering the typing errors caused 
by the use of that abbreviation, the final rule spells out the word 
``liter'' wherever it appears.

Section 71.55  General Requirements for Fissile Material Packages

    One comment suggested that by adding the word ``full'' to the water 
reflection criterion of Sec. 71.55(b)(3), the NRC has added more cost 
with no apparent benefit ``* * * since transport limits already take 
this consideration into account.'' The latter part of this comment 
probably refers to the ``transport index'' controls that limit the 
number of packages which can be transported and stored together, but do 
not consider the safety of an individual package in isolation. Addition 
of the word ``full'' in Sec. 71.55(b)(3) is a matter of clarification. 
NRC has always required ``full'' reflection wherever reflection is 
required. IAEA regulations required ``full'' reflection in the 1973 
edition, and go a step further in the 1985 edition, to define ``full'' 
as ``water 20-cm thick (or its equivalent).'' NRC has retained the word 
``full,'' in Sec. 71.55(b)(3), and has added the word ``full,'' in 
Sec. 71.55(e)(3), for consistency.
    A commenter agrees that the proposed Part 71 begins to simplify the 
system of shipping fissile material but that most of the difficulties 
still exist. The commenter advocates development of ``a system of 
performance-oriented packaging,'' to reduce the current complexity of 
the ``design-oriented package choices.'' NRC agrees that there are a 
number of radiation control design requirements that apply to the 
fissile material packages as well as to packages of other radioactive 
material. However, NRC views the criticality control provisions as 
performance-oriented rather than design-oriented. NRC must specify the 
conditions against which the package must be designed. Without the 
environmental tests and package objectives, there would be no level of 
protection against which to design packages.

Section 71.61  Special Requirement for Irradiated Nuclear Fuel 
Shipments

    One comment recommended that the rule clarify that the deep 
immersion test is to be applied to an otherwise undamaged package. This 
important detail is implied, but not specifically stated. The 
Commission agrees and has made that clarification.
    In the final rule, this section has been modified to require that 
the external pressure test be applied directly to the containment 
system of a package. NRC does not believe the external structure should 
play a part in helping the containment system of a package withstand an 
external pressure test and has chosen to ignore its existence in 
specifying the requirement.
    A comment recommended that the word ``rupture,'' as used in this 
requirement, be defined as a gross structural collapse and not just an 
inleakage of water. Although the word ``rupture'' in the proposed rule 
did mean gross structural collapse, NRC has since decided that the term 
``rupture'' cannot be determined by engineering analysis. NRC has 
decided to change the acceptance criteria for the deep immersion test 
from ``rupture'' to ``collapse, buckling, or inleakage of water.''
    A comment stated that this requirement should include the 1-hour 
time specification included in the IAEA requirement to avoid later 
misinterpretation of the test. The NRC agrees that adding the 1-hour 
test specification would help prevent confusion between IAEA and 
domestic regulations, and has included the time specification. 

[[Page 50257]]

    A comment noted that the term ``at least'' is used two times in the 
proposed requirement, thereby creating an opportunity for 
misinterpretation. Although the term is used in the IAEA text, the NRC 
agrees with the commenter that it serves no useful purpose and has 
deleted the term.
    A comment stated that the deep-water immersion test should be 
clarified to ensure that an engineering evaluation is an acceptable 
alternative to a physical test because an actual 200-m test would be 
costly and difficult. NRC believes it is clear that an engineering 
evaluation is acceptable because the equivalent external gauge pressure 
is specified in the text of the requirement. The provisions of 
Sec. 71.41(a) are intended to allow the use of engineering evaluations 
when they are reasonably applied.
    The remaining three comments relating to this section all deal with 
transition periods and special provisions for casks for which there 
will be no further fabrication and that are not used internationally. 
The earlier portion of this preamble dealing with the provisions of 
Sec. 71.13 presents the NRC view on these matters.

Section 71.63  Special Requirements for Plutonium Shipments

    Four comments argued that the extension of this provision to 
radionuclides other than plutonium is unjustified and that the 
provision, even without the extension to other radionuclides, differs 
from IAEA rules and is inconsistent with the principles of IAEA rules. 
Two of the commenters argued further that the existing provisions, if 
examined in the light of current regulatory analyses, probably could 
not be justified.
    NRC recognizes that some requirements have been added to the 
regulations over the years strictly on the basis of prudent judgment. 
Because the basis for current rules is not a part of this rulemaking 
action, NRC will simply refrain from extending the present rule to 
other radionuclides.
    One commenter argued that the rule should be rewritten using 
multiples of the A2 values, not only to define radionuclides 
subject to the rule, but also to define the level of activity at which 
the extra requirements come into effect. Because the extension to other 
radionuclides is being withdrawn, the inclusion of A values does not 
appear to improve the requirement.

Section 71.71  Normal Conditions of Transport

    Three comments noted that the provision of IAEA's paragraph 528 
requiring consideration of a temperature range from -40  deg.C to +70 
deg.C for the components of the packaging is not reflected in Part 71. 
NRC omitted this provision because NRC does not want to limit the high 
end temperature consideration to 70  deg.C because that would imply 
that +70  deg.C is the highest temperature that has to be considered 
for package design. This does not take into account the considerably 
higher temperatures resulting from decay heat in certain Type B 
packages.
    Three comments noted that 10 CFR 71.71(c)(4) prescribes an 
increased external pressure specification of 140 kPa absolute but IAEA 
regulations do not have that exact requirement. NRC believes there is a 
need for an external pressure test for normal conditions to ensure that 
a package filled at low pressure or high altitude will withstand an 
external pressure increase. The additional pressure test has been 
retained.
    Three comments observed that Sec. 71.71(c)(7) states that the free 
drop test be conducted between 1.5 and 2.5 hours after the conclusion 
of the water spray test but the same requirement is not included in the 
IAEA regulations. The IAEA rules, however, do include restrictions, in 
paragraph 620, on the timing of the mechanical tests after the water 
spray test. NRC has retained the water spray test as is and believes 
the NRC test meets the intent of the IAEA test.
    One comment noted that with the deletion of the fissile classes, 
the corner drop test, which was required only for Fissile Class II 
packages, is proposed to be applied to all fissile packages. The 
commenter argued that for a large and heavy package, such as a spent 
fuel shipping cask, ``it is considered highly implausible for a package 
to undergo a one-foot corner drop as a normal condition of transport. 
Only a free drop with the package in its normal orientation should be 
specified as a normal condition of transport for large and heavy 
packages, therefore saving valuable analysis effort and time.''
    NRC agrees with the comment and has deleted the corner drop test 
for fiberboard, wood, or fissile material rectangular packages weighing 
more than 50 kg (110 lb), and for fissile material cylindrical packages 
weighing more than 100 kg (220 lb). For these packages, NRC does not 
believe that the corner drop tests are significant in developing a safe 
fissile material package.

Section  71.73  Hypothetical Accident Conditions

    One comment stated that reversing the order of the two immersion 
tests in Secs. 71.73 (c)(5) and (c)(6) would restore the order of the 
tests, which must be run consecutively, and would therefore clarify the 
text. NRC agrees and has made the change.
    One comment recommended that the temperature extremes specified for 
the initial test conditions in Sec. 71.73(b) be given a reasonable 
tolerance because ambient air temperatures cannot be controlled. NRC 
agrees that temperatures, as with other required parameters of the test 
conditions, cannot be accurately controlled. NRC's position, however, 
is not to establish tolerances, but to require that the effects of test 
conditions different from those specified be analyzed as part of the 
overall evaluation. Every analysis would then be normalized to the same 
set of specifications.
    One comment recommended that the word ``single,'' in the second 
line of the thermal test in Sec. 71.73(c)(4), should be ``simple''. NRC 
agrees and has made that change.
    Two comments asked that NRC include some information as to how the 
effects of solar radiation should be treated. One comment stated, ``The 
solar insolation can be a significant factor and should be consistently 
evaluated.'' Others have argued that the effects of solar insolation 
are insignificant compared with the thermal effects of the fire test 
and should be ignored.
    NRC adopts the view of the thermal experts who participated in 
developing the IAEA regulations. Those experts thought the effects of 
solar radiation may be neglected before and during the thermal test but 
that such effects should be considered in the subsequent evaluation of 
the package response.
    One comment recommended the development of guidance on how 
designers should interpret the revised thermal test requirement. 
Although there is guidance provided in the IAEA's companion documents 
to its transportation regulations (IAEA Safety Series No. 7, 
``Explanatory Material for the IAEA Regulations for the Safe Transport 
of Radioactive Material--1985 Edition,'' and IAEA Safety Series No. 37, 
``Advisory Material for the IAEA Regulations for the Safe Transport of 
Radioactive Material--1985 Edition''), further guidance may be 
necessary. If so, it is the industry that can best propose guidance, 
based on its capabilities. If coordinated under the auspices of the 
American National Standards Institute (ANSI), Committee N-14, with NRC 
representation, there is a good chance that a consensus standard could 
be developed that could be endorsed by NRC as a satisfactory means to 
satisfy regulatory requirements. 

[[Page 50258]]

    One comment stated that packages that are subjected to the crush 
test should not also be subjected to the 30-foot free drop test, as 
required in the proposed rule. Instead, consistent with IAEA, the crush 
test should be in lieu of the 30-foot free drop test.
    NRC believes that the crush test and the free drop test impart 
different types of loadings onto the package. Having sufficient crush 
resistance for the crush test does not ensure the adequacy of the 
package under the inertial loadings that occur during the 30-foot drop 
tests. NRC believes that it is important for packages to have 
resistance to impact and that the crush test should not be a substitute 
for the impact test.
    One comment stated that a crush scenario is not likely during 
``dedicated'' shipments because heavy loads are not placed above the 
shipment at any time during transport. The comment questioned the 
applicability of the test for dedicated shipments, and requested that 
at least an engineering evaluation be allowed as an alternative to a 
physical test. NRC has made it clear (see Sec. 71.41) that appropriate 
analyses may be used to demonstrate the ability of a package to meet 
crush test conditions.

Section 71.75  Qualifications of Special Form Radioactive Material

    One comment indicates that changes in Sec. 71.75(a) from the 
current rule have changed the concept of special form from being a 
provision for special properties of the radioactive material contents 
of the package to being a provision for special properties of the 
package--a change from qualifying a ``special form source'' to 
qualifying a ``special form package.''
    NRC regrets the confusion, but intended no substantive change to 
the concept of special form. Special form criteria in this final rule 
have been brought closer to those of DOT, but still without any basic 
changes.
    One comment noted that the reference in Sec. 71.75(e) 
[Sec. 71.75(d), in the final rule], to a standard of the International 
Standard Organization (ISO) is vague and should be made more specific.
    Although the ISO standard could be written in all its detail in 
Part 71, rather than simply referenced there, most comments over the 
years have encouraged NRC to have less repetition and more simple 
references to other requirements.

Section 71.83  Assumptions as to Unknown Properties

    One comment pointed out an error in line 7 of Sec. 71.83, where the 
proposed rule referred to ``known properties'', where it should have 
referred to ``unknown properties.'' That error has been corrected.

Section 71.85  Preliminary Determinations

    One comment recommended that the term ``durable'' in the context of 
``durably mark the packaging,'' as in Sec. 71.85, be defined in terms 
of the conditions that the markings on the packaging must be able to 
withstand. When developing its regulations, NRC must decide at what 
level of detail they are to be written. Sometimes that level of detail 
is changed as a result of experience if a widespread misuse of a 
standard becomes known because of a lack of detail. NRC is not aware of 
any problem with the term ``durably,'' even though it has been used 
since 1968 in the preliminary determinations section. In the absence of 
a significant problem, NRC prefers to leave the term as is.

Section 71.87  Routine Determinations

    One comment recommended that NRC's Table V ``Removable External 
Radioactive Contamination Wipe Limits,'' be used by DOT in place of its 
Table 11. NRC notes that the only significant difference between the 
two tables is that the term ``low toxicity alpha emitters'' is replaced 
by its definition in the NRC table. The NRC final rule simply refers to 
the DOT requirement (49 CFR 173.443) for maximum permissible 
contamination limits.

Section 71.88  Air Transport of Plutonium

    One comment recommended that the forward tie-down specification of 
9 g detailed in Sec. 71.88(c)(2) be reduced to 1.5 g for plutonium 
packages transported on a Boeing 747 aircraft. The reason for this 
recommendation has to do with the 14 CFR 25.561 regulatory requirement 
of the Federal Aviation Administration (FAA), that the supporting 
structure of an airplane must be designed to restrain, up to specified 
inertial forces, including 9-g in the forward direction, ``* * * each 
item of mass that could injure an occupant if it came loose in a minor 
crash landing.'' NRC, in prescribing tie-down requirements for 
plutonium packages in aircraft, took note of the supporting structure 
requirements of the FAA and required a 9-g tie-down system for the 
package on the main deck of the aircraft. The Boeing 747 cargo 
aircraft, however, with no passengers and the cockpit located above the 
main deck, is not subject to the requirements of 14 CFR 25.561 because 
there are no occupants to injure if ``* * * the package came loose in a 
minor crash landing.'' Thus, the Boeing 747 ``Weight and Balance 
Manual,'' DG-13700, shows a load factor of 1.5 g in the forward 
direction.
    The purpose of the NRC tie-down requirement was not to protect 
occupants of the aircraft from cargo that has come loose in a minor 
crash landing. Therefore, the comparison with the FAA supporting 
structure requirement is not germane. The purpose of the NRC 
requirement was to protect the plutonium package from the uncontrolled 
potential for damage inherent in having the package unrestrained in a 
crash landing.
    Paragraph (c) of Sec. 71.88 proposed a requirement that the 
licensee make special arrangements with the carrier on where to place 
the plutonium cargo in the aircraft, how to tie it down, and what 
restrictions are to be placed on other cargo. Recognizing that these 
restrictions would be more appropriately placed directly on the carrier 
rather than through the shipper, the DOT has placed these restrictions 
in its air carrier regulations (Sec. 175.704 of 49 CFR Part 175, 
``Carriage By Aircraft.'') These regulations are now referenced in 
Sec. 71.88.

Section 71.95  Reports

    All three public comments on this section were directed at the 
newly proposed provisions of paragraph (c), which require a 30-day 
report of ``* * * instances in which the conditions of approval in the 
certificate of compliance were not observed in making a shipment.''
    One comment requested clarification whether Sec. 71.95(c) applies 
to shippers or receivers.
    The scope of Part 71 (Sec. 71.0(c)) makes the regulation applicable 
only to shippers of radioactive material. Therefore, Sec. 71.95(c) 
applies only to shippers of radioactive material. However, shipment 
deficiency may be detected by the receiver of the shipment. If the 
receiver reports that deficiency to the shipper, the shipper is 
obligated to report it to NRC. Further, note that 10 CFR Part 21, 
``Reporting of Defects and Noncompliance'', is applicable to receiving 
facilities.
    The other two comments dealt with the substance of the event that 
would prompt the report. One suggested the regulation be more specific 
on conditions that would require a report. The second comment suggested 
that the report include the consequences of the deficient shipment such 
as radioactive contamination, a loosened sealing cap, etc.
    Although both of these suggestions have merit, neither has been 

[[Page 50259]]
    incorporated in the final rule. The purpose of the requirement is to 
provide feedback to NRC on quality assurance program effectiveness by 
an indication of the number and type of packaging and other mistakes 
and on the safety significance of those mistakes by an indication of 
the mistake consequences. NRC believes the reporting requirement should 
retain its broad scope. A large number of reports is not expected. NRC 
also believes that individual follow-up is the only reasonable way to 
uncover any procedural deficiency that might cause mistakes.
    One comment questioned whether this type of report is important 
enough to be required within 30 days. NRC judges that the timing is 
about right, and expects the staff's review of submitted reports to be 
completed within a similar time frame.

Section 71.97  Advance Notification of Shipment of Irradiated Reactor 
Fuel and Nuclear Waste

    Of the five comments submitted on this notification requirement, 
two suggested changing the value for the number of curies in 
Sec. 71.97(b)(3)(iii), so it corresponds to the same limit in the 
regulations of DOT and IAEA. That change has been made.
    The other three comments stated that this requirement was not 
clearly expressed. The requirement has been reorganized in the final 
rule, and consists of the following parts:
    1. Paragraph (a) provides a broad general requirement that 
licensees pre-notify governors of States of any shipments of 
radioactive material going to, through, or across the boundary of the 
State;
    2. Paragraph (b) limits the prenotification requirement to certain 
types of shipments. All the conditions of paragraph (b) must be 
satisfied for the prenotification requirement to apply. The licensed 
material must be required to be in a Type B package, limiting the 
requirement to shipments of relatively high potential hazard. The 
shipment must be destined to a disposal site or to a collection point 
for transport to a disposal site, further limiting the requirement to 
waste material. The quantity of radioactive waste in a single package 
must exceed the limits specified in the DOT regulations for highway-
route controlled quantities. Lastly, for irradiated fuel, the quantity 
contained in a single package must be less than that subject to the 
similar advance notification requirement of 10 CFR 73.37(f).
    3. Paragraphs (c), (d), (e) and (f) contain the details for timing, 
information in the notification, revisions, and cancellation.
    One comment noted that from the wording in Sec. 71.97(a), a reader 
would expect to find exceptions in Sec. 71.97(b). The comment notes 
that the provision does not contain exceptions. NRC agrees with this 
comment and has revised Sec. 71.97(a) for clarity.
    One comment questioned the value of proposed Sec. 71.97(b)(4) 
[Sec. 71.97(b) in the final rule] which required that ``* * * the 
quantity of irradiated fuel is less than that subject to advance 
notification requirements of Sec. 73.37(f) of this chapter.'' Paragraph 
73.37(f) refers to a separate part of the Commission's regulations, 10 
CFR Part 73, ``Physical Protection of Plants and Materials,'' and 
imposes an advance notification requirement for irradiated fuel 
shipments similar to the one under discussion. The scope of Part 73 
(see Sec. 73.1(b)(5)) limits its applicability regarding shipments of 
irradiated reactor fuel to ``* * * quantities that in a single shipment 
both exceed 100 grams in net weight of irradiated fuel, exclusive of 
cladding or other structural or packaging material, and have a total 
radiation dose rate in excess of 100 rems per hour at a distance of 3 
feet from any accessible surface without intervening shielding.'' If 
the quantity of irradiated fuel in a shipment exceeded the quantity 
specified in Sec. 73.1(b)(5), the notification would be made under 
Sec. 73.37(f). If not, the notification would be made under Sec. 71.97. 
The proposed provision in Sec. 71.97(b)(4) was intended to prevent 
duplicate notifications for some shipments.
    The final comment on Sec. 71.97 included a clear rewrite of 
Sec. 71.97(b) that has been used in its entirety in the final rule.

Comments on Appendix A

    Five comments supported the inclusion of new radionuclides in Table 
A-1 of Appendix A as useful and justified. Five other comments pointed 
out errors and inconsistencies between NRC and DOT for the A1/
A2 values in Table A-1. These inconsistencies have been corrected 
in the NRC and DOT final rules.
    Three comments recommended a grandfathering provision for the 
continued authority to transport molybdenum (Mo) 99/technetium (Tc) 99m 
generators, in Type A packages, with radioactivity between the current 
A2 value of 20 Ci and the new A2 value of 13.5 Ci for Mo-99. 
The lower A2 value is the result of a new dosimetric model, for 
beta-emitting radionuclides, to address skin contamination. In the 
preamble to the NRC proposed rule, the NRC noted, with respect to the 
changes in the A1 and A2 values:

    Based on our most current knowledge of radioactive material 
shipments in the United States, the economic impacts of these 
changes are not likely to be large. However, any situations where a 
potential exists for significant economic impacts as a result of 
changes in the A1 or A2 values should be brought to the 
NRC's attention in public comments.

NRC agrees that this is a situation where health care in the United 
States could be significantly impacted as a result of forcing the 
larger quantity Mo-99/Tc-99m generators now transported in Type A 
packages into Type B packages. In view of the favorable experience over 
the years with these generators, NRC and DOT will allow the continued 
domestic transportation of generators that contain up to 20 Ci of 
radioactive material in Type A packages.
    Two similar proposals to grandfather the transportation of carbon-
14, phosphorus-32, sulfur-35, and iodine-125 at existing levels were 
not as persuasive and have not been adopted. The decrease in A1 
and A2 values would apparently force many shipments out of the 
``limited quantity'' category, where they are excepted from 
specification packaging, shipping papers and certification, and marking 
and labeling requirements, and into the ``Type A'' category.
    Although there are clearly more packaging and communication 
requirements associated with the ``Type A'' category than with the 
``limited quantity'' category, NRC does not view that change as 
creating the same economic impact as a change from the ``Type A'' to 
the ``Type B'' category.
    One comment suggested that the radionuclides einsteinium-253 and 
einsteinium-254 be added to Table A-1 because shipment of those 
transuranics are increasing in number and the default values are not 
expected to be adequate. NRC has added those radionuclides and will 
also propose them for addition to the IAEA regulations. Until they are 
included in IAEA Safety Series No. 6, however, multilateral approval is 
required for international shipments. This limitation is identified by 
footnote in Table A-1.
    One comment objected to having to obtain NRC approval of A1/
A2 values that are not in Table A-1. In addition to NRC approval, 
international shipments require multilateral approval of A values that 
are not included in the IAEA regulations by each country through or 
into which the consignment is to be transported. The development of A 
values may not be a simple matter, requiring consideration of daughter 

[[Page 50260]]
radionuclides and differing radioactive emissions. Although a competent 
health physicist or nuclear engineer should not have too much 
difficulty determining an A value, NRC must assure that a system exists 
to protect against faulty determinations. Use of the conservative A 
values from Table A-2 does not require regulatory approval.
    One commenter questioned the unlimited values, for A1 and 
A2 in Table A-1, for uranium-235 enriched less than 5 percent. The 
comment argued that U-235 is a fissile material and the unlimited 
values may not be appropriate. The A1/A2 values are for 
radiological, not fissile, considerations. The A1/A2 values 
set the maximum quantity of radioactive material that can be shipped in 
a Type A package (except for LSA); other package characteristics, such 
as heat generation, weight, criticality, external radiation, etc., can 
further limit the quantity of radioactive material in that Type A 
package. Limitations with respect to fissile characteristics, for 
example, are addressed in Secs. 71.53, 71.55, and 71.59. NRC has 
decided to add a clarifying note, currently in the IAEA regulations, to 
the A1/A2 Table in Appendix A of Part 71. The Appendix A note 
reads ``Where values of A1 and A2 are unlimited, it is for 
radiation control purposes only. For nuclear criticality safety, some 
materials are subject to controls placed on fissile material.''
    Finally, one comment suggested that we eliminate the specific 
activity column from Table A-1. The comment argues that ``Specific 
activity information is not required or explained in the regulations, 
and it is difficult to keep the information accurate.''
    Although the NRC is in basic agreement with the comment and would 
have no problem in eliminating the specific activity data from Part 71 
if there were a good source of comparable data available for the times 
it is needed to implement the transportation regulations. NRC is not 
familiar with any good substitute source. Though IAEA Safety Series No. 
37, ``Advisory Material for the IAEA Regulations for the Safe Transport 
of Radioactive Material (1985 Edition),'' third edition, published in 
June 1987, includes a table of half-lives and specific-activities, 
there is no indication yet of a system of periodic reviews that would 
keep that information up to date.

Comments on Draft Regulatory Analysis

    Ten persons commented on the impacts associated with the proposed 
changes to limit the content of LSA/SCO packages to 2A1. The main 
thrust of these comments is that the impacts are much greater than 
presented. In part in response to these comments, NRC has adopted in 
the final rule the IAEA LSA/SCO package limit of 10 mSv/h (1 rem/h) at 
3 m, in lieu of the proposed 2A1 limit.
    Because the NRC data base for determining the additional shipments 
expected to be caused by the proposed rule dated back to 1980, and 
because a clear preference was developing in the public comments for 
the IAEA radiation level limit rather than the 2A1 limit, NRC 
repeated its analysis using more recent data. An NRC contractor 
gathered 1989 data from the 3 shallow land burial facilities for all 
waste shipments of resins, evaporator bottoms, and filter media. The 
contractor analyzed the characteristics of those 4600 Type A cask 
shipments and found that approximately 150 of those shipments would 
have exceeded the IAEA limit. NRC assumes that each shipment exceeding 
the limit is split into 2 shipments due to the smaller capacity of Type 
B packaging. Thus 150 additional shipments are caused by the LSA limit.
    The impacts of preparing additional packages of LSA waste for 
shipment and receiving those additional shipments at the burial ground 
were absent from the draft regulatory analysis. One comment advised the 
NRC of the results of an exposure study which concluded that the extent 
of the collective exposure for preparation and receipt of waste casks 
was approximately 0.5 person-rem per shipment. The NRC noted that half 
of the 0.5 person-rem per shipment factor multiplied by the 4600 waste 
cask shipments per year from the new data base corresponds fairly well 
to a large portion of the 1726 person-rem collective exposure reported 
for all light water reactors for 1986 under the category ``waste 
processing'' by Barbara G. Brooks, NRC, and D. Hagemeyer, SAIC in 
NUREG-0713, Vol. 8, dated August 1989 (this version was current at the 
time the contractor prepared the regulatory analysis). On the basis of 
this data, NRC has accepted the 0.5 man-rem per shipment number as a 
reasonable estimate. Multiplying that 0.5 man-rem per shipment 
conversion factor by the 150 additional shipments which the limit of 1 
rem per hour at 3 meters would cause, the effect of the limit would be 
75 person-rem per year.
    Because the IAEA LSA provisions permit a greater quantity of LSA/
SCO material to be shipped in a package, fewer packages and shipments 
are needed to transport a given quantity of material. The estimated 
burden on industry from the final rule is therefore less than that for 
the proposed rule. The NRC draft regulatory analysis dated November, 
1987 developed industry costs resulting from a 2A1 limit on LSA 
shipments of $1.7 million per year. These costs consist of package 
costs and shipment costs resulting from an estimated 311 additional 
cask shipments per year. Through the same simple modeling used in the 
older analysis, the new NRC regulatory analysis shows increased dollar 
costs associated with the 150 additional LSA/SCO shipments of $1.0 
million per year. These estimates include differential package costs 
and differential shipping and handling costs, annualizing and summing 
each component. These estimates do not include cost components 
recognized but not quantified in the public comments as training, 
procedure revisions, computer program changes and upgrades, insurance 
premiums, and disposal costs.
    There were no significant comments related to the projected number 
of non-radiological deaths and injuries associated with the increased 
shipments caused by the new standards.

Agreement State Compatibility

    Section 274d.(2) of the Atomic Energy Act of 1954, as amended, 
requires that before entering into an agreement with any State, the 
Commission shall make a determination that the State's program is 
compatible with the Commission's program. Section 274g authorizes and 
directs the Commission to cooperate with the States in the formulation 
of standards to assure that State and Commission programs will be 
coordinated and compatible. The basic objective of NRC's State 
Agreements Program has been to achieve uniformity among the various 
programs to the maximum extent practicable recognizing that the States 
must be allowed some flexibility to accommodate local conditions. Under 
this Program, procedures have established criteria for better defining 
compatibility, and for determining the degree to which States 
regulations must show uniformity with Commission regulations. In 
practice, the Commission's regulations are categorized as Division 1-4 
Rules according to the degree of State regulation uniformity required, 
as summarized in the following table:

------------------------------------------------------------------------
   Division              Agreement State regulation uniformity          
------------------------------------------------------------------------
1............  Agreement States are expected to adopt, essentially      
                verbatim, the regulation to provide consistency between 
                Federal and State requirements.                         

[[Page 50261]]
                                                                        
2............  Agreement States have the flexibility to adopt similar or
                more stringent requirements based on their radiation    
                protection experience, professional judgements, and     
                community values.                                       
3............  Agreement States should adopt the requirement, but there 
                is no degree of uniformity between NRC and Agreement    
                States required.                                        
4............  Agreement States should not adopt the requirement since  
                these are regulatory functions reserved to NRC.         
------------------------------------------------------------------------



    The final rule does not affect the current compatibility 
categorization of Part 71 regulations. The following table lists the 
Part 71 Sections and corresponding rule categorization (Division 1-4):

----------------------------------------------------------------------------------------------------------------
         Division                     Section                                     Title                         
----------------------------------------------------------------------------------------------------------------
1.........................  71.4......................  Definitions.                                            
1.........................  71.5......................  Transportation of Licensed Material.                    
1.........................  71.10.....................  Exemption for Low-Level Materials.                      
1.........................  Appendix A................  Determination of A1 and A2.                             
2.........................  71.12.....................  General License: NRC-Approved Package.                  
2.........................  71.13.....................  Previously Approved Package.                            
2.........................  71.14.....................  General License: DOT Specification Container.           
2.........................  71.16.....................  General License: Use of Foreign Approved Package.       
2.........................  71.81.....................  Applicability of Operating Controls and Procedures.     
2.........................  71.85.....................  Preliminary Determinations.                             
2.........................  71.87.....................  Routine Determinations.                                 
2.........................  71.88.....................  Air Transport of Plutonium.                             
2.........................  71.89.....................  Opening Instructions.                                   
2.........................  71.97.....................  Advance Notification of Shipment of Irradiated Reactor  
                                                         Fuel and Nuclear Waste.                                
3.........................  71.0......................  Purpose and Scope.                                      
3.........................  71.1......................  Communications.                                         
3.........................  71.2......................  Interpretations.                                        
3.........................  71.3......................  Requirement for License.                                
3.........................  71.7......................  Completeness and Accuracy of Information.               
3.........................  71.8......................  Specific Exemptions.                                    
3.........................  71.9......................  Exemption of Physicians.                                
3.........................  71.91.....................  Records.                                                
3.........................  71.93.....................  Inspections and Tests.                                  
3.........................  71.95.....................  Reports.                                                
3.........................  71.99.....................  Violations.                                             
3.........................  71.101....................  Quality Assurance Requirements.                         
3.........................  71.103....................  Quality Assurance Organization.                         
3.........................  71.105....................  Quality Assurance Program.                              
3.........................  71.107....................  Package Design Control.                                 
3.........................  71.109....................  Procurement Document Control.                           
3.........................  71.111....................  Instructions, Procedures, and Drawings.                 
3.........................  71.113....................  Document Control.                                       
3.........................  71.115....................  Control of Purchased Material, Equipment, and Services. 
3.........................  71.117....................  Identification and Control of Materials, Parts, and     
                                                         Components.                                            
3.........................  71.119....................  Control of Special Process.                             
3.........................  71.121....................  Internal Inspection.                                    
3.........................  71.123....................  Test Control.                                           
3.........................  71.125....................  Control of Measuring and Test Equipment.                
3.........................  71.127....................  Handling, Storage, and Shipping Control.                
3.........................  71.129....................  Inspection, Test and Operating Status.                  
3.........................  71.131....................  Nonconforming Materials, Parts, or Components.          
3.........................  71.133....................  Corrective Action.                                      
3.........................  71.135....................  Quality Assurance Records.                              
3.........................  71.137....................  Audits.                                                 
4.........................  71.6......................  Information Collection Requirements: OMB Approval.      
4.........................  71.18.....................  General License: Fissile Material, Limited Quantity per 
                                                         Package.                                               
4.........................  71.20.....................  General license: Fissile Material, Limited Moderator per
                                                         Package.                                               
4.........................  71.22.....................  General License: Fissile Material, Limited Quantity,    
                                                         Controlled Shipment.                                   
4.........................  71.24.....................  General License: Fissile Material, Limited Moderator,   
                                                         Controlled Shipment.                                   
4.........................  71.31.....................  Contents of Application.                                
4.........................  71.33.....................  Package Description.                                    
4.........................  71.35.....................  Package Evaluation.                                     
4.........................  71.37.....................  Quality Assurance.                                      
4.........................  71.38.....................  Renewal of a Certificate of Compliance or Quality       
                                                         Assurance Program Approval.                            
4.........................  71.39.....................  Requirement for Additional Information.                 
4.........................  71.41.....................  Demonstration of Compliance.                            
4.........................  71.43.....................  General Standards for all Packages.                     
4.........................  71.45.....................  Lifting and Tie-down Standards for all Packages.        
4.........................  71.47.....................  External Radiation Standards for all Packages.          
4.........................  71.51.....................  Additional Requirements for Type B Packages.            

[[Page 50262]]
                                                                                                                
4.........................  71.52.....................  Exemption for Low-Specific-Activity (LSA) Packages.     
4.........................  71.53.....................  Fissile Material Exemptions.                            
4.........................  71.55.....................  General Requirements for Fissile Material Packages.     
4.........................  71.59.....................  Standards for Arrays of fissile Material Packages.      
4.........................  71.61.....................  Special Requirement for Irradiated Nuclear Fuel         
                                                         Shipments.                                             
4.........................  71.63.....................  Special Requirements for Plutonium Shipments.           
4.........................  71.64.....................  Special Requirements for Plutonium Air Shipments.       
4.........................  71.65.....................  Additional Requirements.                                
4.........................  71.71.....................  Normal Conditions of Transport.                         
4.........................  71.73.....................  Hypothetical Accident Conditions.                       
4.........................  71.74.....................  Accident Conditions for Air Transport of Plutonium.     
4.........................  71.75.....................  Qualification of Special Form Radioactive Material.     
4.........................  71.77.....................  Qualification of LSA-III Material.                      
4.........................  71.83.....................  Assumptions as to Unknown Properties.                   
4.........................  71.100....................  Criminal Penalties.                                     
----------------------------------------------------------------------------------------------------------------



Petitions for Rulemaking

    Three petitions for rulemaking were filed with the NRC in 
connection with the rules for transporting LSA radioactive material. 
The substance of each of the three petitions was essentially the same, 
to request that NRC exempt LSA materials from its requirements in Part 
71.
    The petitioners were the Energy Research and Development 
Administration (now the U.S. Department of Energy) in its letter dated 
July 23, 1975 (PRM-71-1); ANSI Committee N14, in its letter dated March 
10, 1976 (PRM-71-2); and Chem-Nuclear Systems, Inc., in its letter 
dated November 22, 1976 (PRM-71-4). At the time these petitions were 
filed, DOT regulated carriers and shippers of small quantities of all 
radioactive materials (including LSA materials) through provisions in 
its regulations in 49 CFR Parts 170-189, whereas NRC regulated shippers 
of fissile material and of larger quantities of other radioactive 
materials (including LSA materials) through its regulations in Part 71 
and its licensing program. All three petitioners argued that the 
control NRC was exerting over transportation of LSA materials created 
an inconsistency between NRC regulations and those of the IAEA and 
should be discontinued. A proposed rule that would have provided the 
exemption for LSA materials requested in the petitions was published by 
NRC for public comment on August 17, 1979 (44 FR 48234). Before 
finalization of that rule, however, a deficiency in the new LSA 
requirements, as proposed, was recognized so that the entire LSA 
proposal, including the exemption, was withdrawn. In the interim, the 
corresponding deficiency in the LSA requirements in the IAEA 
regulations was recognized and corrected. That correction is discussed 
under the ``major modifications from proposed rule'' section of this 
preamble. This correction is implemented in both DOT regulations and 
NRC regulations.
    The exemption requested in the three petitions has been superseded 
by the changes in LSA requirements. The LSA requirements imposed in NRC 
regulations are an integral part of the NRC/DOT regulatory scheme for 
LSA materials. This scheme is based on IAEA regulations. There is an 
exemption provided for LSA materials in Sec. 71.10 that clearly defines 
the level where NRC regulations impose additional packaging 
requirements. For the above reasons, NRC has denied the petitions.

Administrative Correction

    At about the same time the Notice of Proposed Rulemaking regarding 
compatibility with IAEA transportation regulations was published for 
public comment on June 8, 1988 (53 FR 21550), a separate notice of 
final rulemaking was issued, by NRC, affecting the retention periods 
for records (53 FR 19240, May 27, 1988). Included in that separate 
notice were changes to the transportation regulations in Part 71, 
specifically to Secs. 71.105, ``Quality assurance program,'' and 
71.135, ``Quality assurance records.'' Because the two rules were being 
processed at the same time by different organizations, NRC's internal 
controls failed to recognize that the new quality assurance provisions 
needed to be incorporated in the June 8, 1988, notice of proposed 
rulemaking. No written comments were filed with respect to the quality 
assurance sections proposed, although two phone calls were received 
advising NRC of its error. The quality assurance changes that were made 
effective by the final rule, published on May 27, 1988, are included in 
this final rule.

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, that this rule is not a major Federal 
action significantly affecting the quality of the human environment, 
and therefore an environmental impact statement (EIS) is not required.
    The Commission's ``Final Environmental Statement on the 
Transportation of Radioactive Material by Air and Other Modes,'' NUREG-
0170,3 dated December 1977, is NRC's generic EIS, covering all 
types of radioactive material transportation by all modes (road, rail, 
air, and water). From the Commission's latest survey of radioactive 
material shipments and their characteristics, ``Transport of 
Radioactive Material in the United States,'' SAND 84-7174, April 1985, 
it can be concluded that current radioactive material shipments are not 
so different from those evaluated in NUREG-0170 as to invalidate the 
results or conclusions of that EIS. Environmental impacts associated 
with this rulemaking are evaluated in ``Regulatory Analysis of Changes 
to 10 CFR Part 71--NRC Regulations on Packaging and Transportation of 
Radioactive Material,'' dated April 1995.

    \3\ Copies of NUREG-0170 may be purchased from the 
Superintendent of Documents, U.S. Government Printing Office, P.O. 
Box 37082, Washington, DC 20013-7082. Copies are also available from 
the National Technical Information Service, 5285 Port Royal Road, 
Springfield, VA 22161. A copy is also available for inspection and 
copying for a fee in the NRC Public Document Room, 2120 L Street, 
NW. (Lower Level), Washington, DC.
---------------------------------------------------------------------------

    NUREG-0170 established the non-accident related radiation exposures 
associated with transportation of radioactive material in the United 
States as 98 person-Sv (9800 person-rem) which, based on the 
conservative linear 

[[Page 50263]]
radiation dose hypothesis, resulted in a maximum of 1.7 genetic effects 
and 1.2 latent cancer effects per year. More than half this impact 
resulted from shipment of medical-use radioactive materials. Accident 
related impacts were established at a maximum of one genetic effect and 
one latent cancer fatality for 200 years of transporting radioactive 
materials. The principal nonradiological impacts were found to be two 
injuries per year, and less than one accidental death per 4 years. In 
contrast, non-accident related radiation exposures associated with this 
rulemaking would be increased by 0.75 person-Sv/y (75.0 person-rem/y), 
whereas accident related impacts would be decreased by approximately 
0.006 person-Sv/y (0.6 person-rem/y). Nonradiological traffic injuries 
would be increased by 0.06 per year and nonradiological traffic deaths 
by 0.003 per year (less than 1 accidental death per 330 years). These 
impacts are judged to be insignificant compared with the baseline 
impacts established in NUREG-0170.
    The environmental assessment and finding of no significant impact 
on which this determination is based are available, for inspection, at 
the NRC Public Document Room, 2120 L Street NW. (Lower Level), 
Washington, DC. Single copies of the environmental assessment and 
finding of no significant impact are also available from the contact 
listed under the Addresses heading.

Paperwork Reduction Act Statement

    This final rule amends information collection requirements that are 
subject to the Paperwork Reduction Act of 1980 (44 U.S.C. 3501 et 
seq.). These requirements were approved by the Office of Management and 
Budget, Approval Number 3150-0008.
    The public reporting burden for this collection of information is 
estimated to average 7 hours per response, including the time for 
reviewing instructions, searching existing data sources, gathering and 
maintaining the data needed, and completing and reviewing the 
collection of information. Send comments regarding this burden estimate 
or any other aspect of this collection of information, including 
suggestions for reducing this burden, to the Information and Records 
Management Branch (T-6F33), U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001; and to the Desk Officer, Office of 
Information and Regulatory Affairs, NEOB-10202, (3150-0008), Office of 
Management and Budget, Washington, D.C. 20503.

Regulatory Analysis

    The NRC has prepared a regulatory analysis on this final 
regulation. The analysis examines the costs and benefits of the 
alternatives considered by NRC. Interested persons may examine a copy 
of the regulatory analysis at the NRC Public Document Room at 2120 L 
Street NW. (Lower Level), Washington, DC. Single copies of the analysis 
may be obtained from the contact listed under the Addresses heading.

Regulatory Flexibility Act Certification

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

Backfit Analysis

    The Commission has determined that the backfit rule does not apply 
to the Part 71 final rule because the final rule is not a backfit under 
10 CFR Part 50.109. However, NRC analyzed the accident-resistant 
packaging requirement for the specified LSA shipments and found that 
there is an increase in overall protection to be derived from the 
requirement and that direct and indirect costs of implementation are 
justified in view of this increased protection.
    The factors normally considered in a backfit analysis are evaluated 
in the ``Regulatory Analysis of Changes to 10 CFR Part 71--NRC 
Regulations on Packaging and Transportation of Radioactive Material,'' 
dated April 1995. That evaluation shows very small changes in accident 
risks as a result of the adoption of the revision, but some reduction 
in maximum consequences given an accident. The evaluation shows broad 
improvement in NRC regulatory consistency with IAEA, at an initial cost 
of $1.375 million to industry, and continual annual costs to industry 
of $1.0 million (See Table S.1 of Regulatory Analysis). NRC costs are 
estimated at $0.463 million.
    The continuing costs are associated with the addition of new limits 
on the quantity of LSA radioactive material allowed in a single 
transportation package. Internationally, a new limit is considered to 
be a necessary safety requirement to limit the consequences of a severe 
transportation accident involving LSA material.
    The one-time costs are chiefly associated with industry upgrading 
of its package safety analyses to include the proposed new accident 
crush and immersion tests and with NRC review of those new analyses. 
The estimated costs are overstated because of the assumption that all 
licensees using packages approved under earlier regulatory standards 
would take immediate steps to upgrade the package analyses so the 
package approvals would reflect approval, under the latest revised 
standards. Although that is a prudent assumption, absent any reasonable 
basis for predicting actual licensee reaction, there is little reason 
licensees would take any immediate action to upgrade their package 
approvals. Both domestic and international regulations are based on the 
responsible agency's confidence that packages built to a design 
approved under earlier standards are adequately safe for continued use, 
although new package construction to that design would be limited, and 
international use requires approval by all countries through which the 
package is to be transported. In actual practice, some package 
approvals would never be upgraded. Those that would be upgraded would 
be done over a period of several years as guidance and experience in 
upgrading become available.
    Although the regulatory analysis shows a small reduction in 
accident risks from the amendments to this rule and some reduction in 
maximum consequences given an accident, the primary benefit of this 
rulemaking is to achieve consistency in radioactive material 
transportation regulations between the United States and the rest of 
the world. This consistency would not only facilitate the free movement 
of radioactive materials between countries for medical, research, 
industrial, and nuclear fuel cycle purposes, but it would also 
contribute to safety by concentrating the efforts of the world's 
experts on a single set of safety standards and guidance (those of the 
IAEA) from which individual countries could develop their domestic 
regulations. In addition, the accident experience of every country that 
bases its domestic regulations on those of the IAEA could be applied to 
every other country with consistent regulations to improve its safety 
program.
    In summary, the effort to make U.S. regulations compatible with 
those of the IAEA provides major benefits including 

[[Page 50264]]
a substantial increase in the overall protection of the public health 
and safety, and it is associated with short-term and relatively minor 
costs that are justified in view of this increased protection. This 
effort is associated with ongoing costs, but the new limit is 
considered to be a justified safety requirement, to limit the 
consequences of a severe transportation accident involving LSA 
material.

List of Subjects in 10 CFR Part 71

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

    For the reasons set out in the preamble and under the authority of 
the Atomic Energy Act of 1954, as amended, the Energy Reorganization 
Act of 1974, as amended, and 5 U.S.C. 552 and 553, 10 CFR part 71 is 
revised to read as follows:

PART 71--PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

Subpart A--General Provisions

Sec.
71.0  Purpose and scope.
71.1  Communications and records.
71.2  Interpretations.
71.3  Requirement for license.
71.4  Definitions.
71.5  Transportation of licensed material.

Subpart B--Exemptions

71.6  Information collection requirements: OMB approval.
71.7  Completeness and accuracy of information.
71.8  Specific exemptions.
71.9  Exemption of physicians.
71.10  Exemption for low-level materials.
71.11  [Reserved]

Subpart C--General Licenses

71.12  General license: NRC-approved package.
71.13  Previously approved package.
71.14  General license: DOT specification container.
71.16  General license: Use of foreign approved package.
71.18  General license: Fissile material, limited quantity per 
package.
71.20  General license: Fissile material, limited moderator per 
package.
71.22  General license: Fissile material, limited quantity, 
controlled shipment.
71.24  General license: Fissile material, limited moderator, 
controlled shipment.

Subpart D--Application for Package Approval

71.31  Contents of application.
71.33  Package description.
71.35  Package evaluation.
71.37  Quality assurance.
71.38  Renewal of a certificate of compliance or quality assurance 
program approval.
71.39  Requirement for additional information.

Subpart E--Package Approval Standards

71.41  Demonstration of compliance.
71.43  General standards for all packages.
71.45  Lifting and tie-down standards for all packages.
71.47  External radiation standards for all packages.
71.51  Additional requirements for Type B packages.
71.52  Exemption for low-specific-activity (LSA) packages.
71.53  Fissile material exemptions.
71.55  General requirements for fissile material packages.
71.57  [Reserved]
71.59  Standards for arrays of fissile material packages.
71.61  Special requirement for irradiated nuclear fuel shipments.
71.63  Special requirements for plutonium shipments.
71.64  Special requirements for plutonium air shipments.
71.65  Additional requirements.

Subpart F--Package, Special Form, and LSA-III Tests

71.71  Normal conditions of transport.
71.73  Hypothetical accident conditions.
71.74  Accident conditions for air transport of plutonium.
71.75  Qualification of special form radioactive material.
71.77  Qualification of LSA-III Material

Subpart G--Operating Controls and Procedures

71.81  Applicability of operating controls and procedures.
71.83  Assumptions as to unknown properties.
71.85  Preliminary determinations.
71.87  Routine determinations.
71.88  Air transport of plutonium.
71.89  Opening instructions.
71.91  Records.
71.93  Inspection and tests.
71.95  Reports.
71.97  Advance notification of shipment of irradiated reactor fuel 
and nuclear waste.
71.99  Violations.
71.100  Criminal penalties.

Subpart H--Quality Assurance

71.101  Quality assurance requirements.
71.103  Quality assurance organization.
71.105  Quality assurance program.
71.107  Package design control.
71.109  Procurement document control.
71.111  Instructions, procedures, and drawings.
71.113  Document control.
71.115  Control of purchased material, equipment, and services.
71.117  Identification and control of materials, parts, and 
components.
71.119  Control of special processes.
71.121  Internal inspection.
71.123  Test control.
71.125  Control of measuring and test equipment.
71.127  Handling, storage, and shipping control.
71.129  Inspection, test, and operating status.
71.131  Nonconforming materials, parts, or components.
71.133  Corrective action.
71.135  Quality assurance records.
71.137  Audits.

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

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

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

Subpart A--General Provisions


Sec. 71.0  Purpose and scope.

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

    \1\ Postal Service Manual (Domestic Mail Manual), section 124.3, 
which is incorporated by reference at 39 CFR 111.1.
---------------------------------------------------------------------------

    (c) The regulations in this part apply to any licensee authorized 
by specific or general license issued by the Commission to receive, 
possess, use, or transfer licensed material, if the licensee delivers 
that material to a carrier for transport, transports the material 
outside the site of usage as specified in the NRC license, or 
transports that material on public highways. No provision of this part 
authorizes possession of licensed material.
    (d) Exemptions from the requirement for license in Sec. 71.3 are 
specified in Sec. 71.10. General licenses for which no NRC package 
approval is required are issued in Secs. 71.14 through 71.24. The 
general license in Sec. 71.12 requires that an NRC certificate of 
compliance or other package approval be issued for the package to be 
used under the general license. Application for package 

[[Page 50265]]
approval must be completed in accordance with subpart D of this part, 
demonstrating that the design of the package to be used satisfies the 
package approval standards contained in subpart E of this part, as 
related to the tests of subpart F of this part. The transport of 
licensed material or delivery of licensed material to a carrier for 
transport is subject to the operating controls and procedures 
requirements of subpart G of this part, to the quality assurance 
requirements of subpart H of this part, and to the general provisions 
of subpart A of this part, including DOT regulations referenced in 
Sec. 71.5.
    (e) The regulations in this part apply to any person required to 
obtain a certificate of compliance or an approved compliance plan 
pursuant to part 76 of this chapter if the person delivers radioactive 
material to a common or contract carrier for transport or transports 
the material outside the confines of the person's plant or other 
authorized place of use.


Sec. 71.1  Communications and records.

    (a) All communications concerning the regulations in this part 
should be addressed to the Director, Office of Nuclear Material Safety 
and Safeguards, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, or may be delivered in person, at the Commission offices, 
at 11545 Rockville Pike, Rockville, Maryland.
    (b) Each record required by this part must be legible throughout 
the retention period specified by each Commission regulation. The 
record may be the original or a reproduced copy or a microform provided 
that the copy or microform is authenticated by authorized personnel and 
that the microform is capable of producing a clear copy throughout the 
required retention period. The record may also be stored in electronic 
media with the capability for producing legible, accurate, and complete 
records during the required retention period. Records such as letters, 
drawings, specifications, must include all pertinent information such 
as stamps, initials, and signatures. The licensee shall maintain 
adequate safeguards against tampering with and loss of records.


Sec. 71.2  Interpretations.

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


Sec. 71.3  Requirement for license.

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


Sec. 71.4  Definitions.

    The following terms are as defined here for the purpose of this 
part. To ensure compatibility with international transportation 
standards, all limits in this part are given in terms of dual units: 
The International System of Units (SI) followed or preceded by U.S. 
standard or customary units. The U.S. customary units are not exact 
equivalents, but are rounded to a convenient value, providing a 
functionally equivalent unit. For the purpose of this part, either unit 
may be used.
    A1 means the maximum activity of special form radioactive 
material permitted in a Type A package. A2 means the maximum 
activity of radioactive material, other than special form, LSA and SCO 
material, permitted in a Type A package. These values are either listed 
in Appendix A of this part, Table A-1, or may be derived in accordance 
with the procedure prescribed in Appendix A of this part.
    Carrier means a person engaged in the transportation of passengers 
or property by land or water as a common, contract, or private carrier, 
or by civil aircraft.
    Certificate holder means a person who has been issued a certificate 
of compliance or other package approval by the Commission.
    Close reflection by water means immediate contact by water of 
sufficient thickness for maximum reflection of neutrons.
    Containment system means the assembly of components of the 
packaging intended to retain the radioactive material during transport.
    Conveyance means:
    (1) For transport by public highway or rail any transport vehicle 
or large freight container;
    (2) For transport by water any vessel, or any hold, compartment, or 
defined deck area of a vessel including any transport vehicle on board 
the vessel; and
    (3) For transport by aircraft any aircraft.
    Exclusive use means the sole use by a single consignor of a 
conveyance for which all initial, intermediate, and final loading and 
unloading are carried out in accordance with the direction of the 
consignor or consignee. The consignor and the carrier must ensure that 
any loading or unloading is performed by personnel having radiological 
training and resources appropriate for safe handling of the 
consignment. The consignor must issue specific instructions, in 
writing, for maintenance of exclusive use shipment controls, and 
include them with the shipping paper information provided to the 
carrier by the consignor.
    Fissile material means plutonium-238, plutonium-239, plutonium-241, 
uranium-233, uranium-235, or any combination of these radionuclides. 
Unirradiated natural uranium and depleted uranium, and natural uranium 
or depleted uranium that has been irradiated in thermal reactors only 
are not included in this definition. Certain exclusions from fissile 
material controls are provided in Sec. 71.53.
    Licensed material means by-product, source, or special nuclear 
material received, possessed, used, or transferred under a general or 
specific license issued by the Commission pursuant to the regulations 
in this chapter.
    Low Specific Activity (LSA) material means radioactive material 
with limited specific activity that satisfies the descriptions and 
limits set forth below. Shielding materials surrounding the LSA 
material may not be considered in determining the estimated average 
specific activity of the package contents. LSA material must be in one 
of three groups:
    (1) LSA-I.
    (i) Ores containing only naturally occurring radionuclides (e.g., 
uranium, thorium) and uranium or thorium concentrates of such ores; or
    (ii) Solid unirradiated natural uranium or depleted uranium or 
natural thorium or their solid or liquid compounds or mixtures; or
    (iii) Radioactive material, other than fissile material, for which 
the A2 value is unlimited; or
    (iv) Mill tailings, contaminated earth, concrete, rubble, other 
debris, and activated material in which the radioactive material is 
essentially uniformly distributed, and the average specific activity 
does not exceed 10-6 A2/g.
    (2) LSA-II.
    (i) Water with tritium concentration up to 0.8 TBq/liter (20.0 Ci/
liter); or
    (ii) Material in which the radioactive material is essentially 
uniformly distributed, 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) in which:
    (i) The radioactive material is essentially uniformly distributed 

[[Page 50266]]
    throughout a solid or a collection of solid objects, or is essentially 
uniformly distributed in a solid compact binding agent (such as 
concrete, bitumen, ceramic, etc.);
    (ii) The radioactive material is relatively insoluble, or it is 
intrinsically contained in a relatively insoluble material, so that, 
even under loss of packaging, the loss of radioactive material per 
package by leaching, when placed in water for 7 days, would not exceed 
0.1 A2; and
    (iii) The average specific activity of the solid does not exceed 2 
x 10-3
A2/g.
    Low toxicity alpha emitters means natural uranium, depleted 
uranium, natural thorium; uranium-235, uranium-238, thorium-232, 
thorium-228 or thorium-230 when contained in ores or physical or 
chemical concentrates or tailings; or alpha emitters with a half-life 
of less than 10 days.
    Maximum normal operating pressure means the maximum gauge pressure 
that would develop in the containment system in a period of 1 year 
under the heat condition specified in Sec. 71.71(c)(1), in the absence 
of venting, external cooling by an ancillary system, or operational 
controls during transport.
    Natural thorium means thorium with the naturally occurring 
distribution of thorium isotopes (essentially 100 weight percent 
thorium-232).
    Normal form radioactive material means radioactive material that 
has not been demonstrated to qualify as ``special form radioactive 
material.''
    Optimum interspersed hydrogenous moderation means the presence of 
hydrogenous material between packages to such an extent that the 
maximum nuclear reactivity results.
    Package means the packaging together with its radioactive contents 
as presented for transport.
    (1) Fissile material package means a fissile material packaging 
together with its fissile material contents.
    (2) Type B package means a Type B packaging together with its 
radioactive contents. On approval, a Type B package design is 
designated by NRC as B(U) unless the package has a maximum normal 
operating pressure of more than 700 kPa (100 lb/in2) gauge or a 
pressure relief device that would allow the release of radioactive 
material to the environment under the tests specified in Sec. 71.73 
(hypothetical accident conditions), in which case it will receive a 
designation B(M). B(U) refers to the need for unilateral approval of 
international shipments; B(M) refers to the need for multilateral 
approval of international shipments. There is no distinction made in 
how packages with these designations may be used in domestic 
transportation. To determine their distinction for international 
transportation, see DOT regulations in 49 CFR Part 173. A Type B 
package approved before September 6, 1983, was designated only as Type 
B. Limitations on its use are specified in Sec. 71.13.
    Packaging means the assembly of components necessary to ensure 
compliance with the packaging requirements of this part. It may consist 
of one or more receptacles, absorbent materials, spacing structures, 
thermal insulation, radiation shielding, and devices for cooling or 
absorbing mechanical shocks. The vehicle, tie-down system, and 
auxiliary equipment may be designated as part of the packaging.
    Special form radioactive material means radioactive material that 
satisfies the following conditions:
    (1) It is either a single solid piece or is contained in a sealed 
capsule that can be opened only by destroying the capsule;
    (2) The piece or capsule has at least one dimension not less than 5 
mm (0.2 in); and
    (3) It satisfies the requirements of Sec. 71.75. A special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on June 30, 1983, (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before July 1, 1985, and a special form 
encapsulation designed in accordance with the requirements of Sec. 71.4 
in effect on March 31, 1996, (see 10 CFR part 71, revised as of January 
1, 1983), and constructed before April 1, 1998, may continue to be 
used. Any other special form encapsulation must meet the specifications 
of this definition.
    Specific activity of a radionuclide means the radioactivity of the 
radionuclide per unit mass of that nuclide. The specific activity of a 
material in which the radionuclide is essentially uniformly distributed 
is the radioactivity per unit mass of the material.
    State means a State of the United States, the District of Columbia, 
the Commonwealth of Puerto Rico, the Virgin Islands, Guam, American 
Samoa, and the Commonwealth of the Northern Mariana Islands.
    Surface Contaminated Object (SCO) means a solid object that is not 
itself classed as radioactive material, but which has radioactive 
material distributed on any of its surfaces. SCO must be in one of two 
groups with surface activity not exceeding the following limits:
    (1) SCO-I: A solid object on which:
    (i) The non-fixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
cm2) does not exceed 4 Bq/cm2 (10-4 microcurie/cm2) 
for beta and gamma and low toxicity alpha emitters, or 0.4 Bq/cm2 
(10-5 microcurie/cm2) for all other alpha emitters;
    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm2 (or the area of the surface if less than 300 
cm2) does not exceed 4x104 Bq/cm2 (1.0 microcurie/
cm2) for beta and gamma and low toxicity alpha emitters, or 
4x103 Bq/cm2 (0.1 microcurie/cm2) for all other alpha 
emitters; and
    (iii) The non-fixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm2 (or the area of the 
surface if less than 300 cm2) does not exceed 4x104 Bq/
cm2 (1 microcurie/cm2) for beta and gamma and low toxicity 
alpha emitters, or 4x103 Bq/cm2 (0.1 microcurie/cm2) for 
all other alpha emitters.
    (2) SCO-II: A solid object on which the limits for SCO-I are 
exceeded and on which:
    (i) The non-fixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 400 Bq/cm\2\ (10-2 microcurie/cm\2\) for beta and gamma 
and low toxicity alpha emitters or 40 Bq/cm\2\ (10-3 microcurie/
cm\2\) for all other alpha emitters;
    (ii) The fixed contamination on the accessible surface averaged 
over 300 cm\2\ (or the area of the surface if less than 300 cm\2\) does 
not exceed 8 x 10\5\ Bq/cm\2\ (20 microcuries/cm\2\) for beta and gamma 
and low toxicity alpha emitters, or 8 x 10 \4\ Bq/cm\2\ (2 microcuries/
cm\2\) for all other alpha emitters; and
    (iii) The non-fixed contamination plus the fixed contamination on 
the inaccessible surface averaged over 300 cm\2\ (or the area of the 
surface if less than 300 cm\2\) does not exceed 8 x 10\5\ Bq/cm\2\ (20 
microcuries/cm\2\) for beta and gamma and low toxicity alpha emitters, 
or 8 x 10\4\ Bq/cm\2\ (2 microcuries/cm\2\) for all other alpha 
emitters.
    Transport index means the dimensionless number (rounded up to the 
next tenth) placed on the label of a package, to designate the degree 
of control to be exercised by the carrier during transportation. The 
transport index is determined as follows:
    (1) For non-fissile material packages, the number determined by 
multiplying the maximum radiation level in millisievert (mSv) per hour 
at one meter (3.3 ft) from the external surface of the package by 100 
(equivalent to the 

[[Page 50267]]
maximum radiation level in millirem per hour at one meter (3.3 ft)); or
    (2) For fissile material packages, the number determined by 
multiplying the maximum radiation level in millisievert per hour at one 
meter (3.3 ft) from the external surface of the package by 100 
(equivalent to the maximum radiation level in millirem per hour at one 
meter (3.3 ft)), or, for criticality control purposes, the number 
obtained as described in Sec. 71.59, whichever is larger.
    Type A quantity means a quantity of radioactive material, the 
aggregate radioactivity of which does not exceed A1 for special 
form radioactive material, or A2, for normal form radioactive 
material, where A1 and A2 are given in Table A-1 of this 
part, or may be determined by procedures described in Appendix A of 
this part.
    Type B quantity means a quantity of radioactive material greater 
than a Type A quantity.
    Uranium--natural, depleted, enriched
    (1) Natural uranium means uranium with the naturally occurring 
distribution of uranium isotopes (approximately 0.711 weight percent 
uranium-235, and the remainder by weight essentially uranium-238).
    (2) Depleted uranium means uranium containing less uranium-235 than 
the naturally occurring distribution of uranium isotopes.
    (3) Enriched uranium means uranium containing more uranium-235 than 
the naturally occurring distribution of uranium isotopes.


Sec. 71.5  Transportation of licensed material.

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

Subpart B--Exemptions


Sec. 71.6  Information collection requirements: OMB approval.

    (a) The Nuclear Regulatory Commission has submitted the information 
collection requirements contained in this part to the Office of 
Management and Budget (OMB) for approval, as required by the Paperwork 
Reduction Act of 1980 (44 U.S.C. 3501 et seq.). OMB has approved the 
information collection requirements contained in this part, under 
control number 3150-0008.
    (b) The approved information collection requirements contained in 
this part appear in Secs. 71.5, 71.6a, 71.7, 71.12, 71.13, 71.31, 
71.33, 71.35, 71.37, 71.38, 71.39, 71.47, 71.85, 71.87, 71.89, 71.91, 
71.93, 71.95, 71.97, 71.101, 71.103, 71.105, 71.107, 71.109, 71.111, 
71.113, 71.115, 71.117, 71.119, 71.121, 71.123, 71.125, 71.127, 71.129, 
71.131, 71.133, 71.135, and 71.137.


Sec. 71.7  Completeness and accuracy of information.

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


Sec. 71.8  Specific exemptions.

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


Sec. 71.9  Exemption of physicians.

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


Sec. 71.10  Exemption for low-level materials.

    (a) A licensee is exempt from all requirements of this part with 
respect to shipment or carriage of a package containing radioactive 
material having a specific activity not greater than 70 Bq/g (0.002 
Ci/g).
    (b) A licensee is exempt from all requirements of this part, other 
than Sec. 71.5 and Sec. 71.88, with respect to shipment or carriage of 
the following packages, provided the packages contain no fissile 
material, or the fissile material exemption standards of Sec. 71.53 are 
satisfied:
    (1) A package containing no more than a Type A quantity of 
radioactive material;
    (2) A package in which the only radioactive material is low 
specific activity (LSA) material or surface contaminated objects (SCO), 
provided the external radiation level at 3 m from the unshielded 
material or objects does not exceed 10 mSv/h (1 rem/h); or

[[Page 50268]]

    (3) A package transported within locations within the United States 
which contains only americium or plutonium in special form with an 
aggregate radioactivity not to exceed 20 curies.
    (c) A licensee is exempt from all requirements of this part, other 
than Secs. 71.5 and 71.88, with respect to shipment or carriage of low-
specific-activity (LSA) material in group LSA-I, or surface 
contaminated objects (SCOs) in group SCO-I.


Sec. 71.11  [Reserved]

Subpart C--General Licenses


Sec. 71.12  General license: NRC-approved package.

    (a) A general license is hereby issued to any licensee of the 
Commission to transport, or to deliver to a carrier for transport, 
licensed material in a package for which a license, certificate of 
compliance, or other approval has been issued by the NRC.
    (b) This general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only to a licensee who--
    (1) Has a copy of the certificate of compliance, or other approval 
of the package, and has the drawings and other documents referenced in 
the approval relating to the use and maintenance of the packaging and 
to the actions to be taken before shipment;
    (2) Complies with the terms and conditions of the license, 
certificate, or other approval, as applicable, and the applicable 
requirements of subparts A, G, and H of this part; and
    (3) Submits in writing to the Director, Office of Nuclear Material 
Safety and Safeguards, U.S. Nuclear Regulatory Commission, Washington, 
DC 20555-0001, before the licensee's first use of the package, the 
licensee's name and license number and the package identification 
number specified in the package approval.
    (d) This general license applies only when the package approval 
authorizes use of the package under this general license.
    (e) For a Type B or fissile material package, the design of which 
was approved by NRC before April 1, 1996, the general license is 
subject to the additional restrictions of Sec. 71.13.


Sec. 71.13  Previously approved package.

    (a) A Type B package previously approved by NRC but not designated 
as B(U) or B(M) in the identification number of the NRC Certificate of 
Compliance, may be used under the general license of Sec. 71.12 with 
the following additional conditions:
    (1) Fabrication of the packaging was satisfactorily completed by 
August 31, 1986, as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A package used for a shipment to a location outside the United 
States is subject to multilateral approval, as defined in DOT 
regulations at 49 CFR 173.403; and
    (3) A serial number that uniquely identifies each packaging which 
conforms to the approved design is assigned to, and legibly and durably 
marked on, the outside of each packaging.
    (b) A Type B(U) package, a Type B(M) package, a low specific 
activity (LSA) material package or a fissile material package, 
previously approved by the NRC but without the designation ``-85'' in 
the identification number of the NRC Certificate of Compliance, may be 
used under the general license of Sec. 71.12 with the following 
additional conditions:
    (1) Fabrication of the package is satisfactorily completed by April 
1, 1999 as demonstrated by application of its model number in 
accordance with Sec. 71.85(c);
    (2) A package used for a shipment to a location outside the United 
States is subject to multilateral approval as defined in DOT 
regulations at 49 CFR 173.403; and
    (3) A serial number which uniquely identifies each packaging which 
conforms to the approved design is assigned to and legibly and durably 
marked on the outside of each packaging.
    (c) NRC will approve modifications to the design and authorized 
contents of a Type B package, or a fissile material package, previously 
approved by NRC, provided--
    (1) The modifications of a Type B package are not significant with 
respect to the design, operating characteristics, or safe performance 
of the containment system, when the package is subjected to the tests 
specified in Secs. 71.71 and 71.73;
    (2) The modifications of a fissile material package are not 
significant, with respect to the prevention of criticality, when the 
package is subjected to the tests specified in Secs. 71.71 and 71.73; 
and
    (3) The modifications to the package satisfy the requirements of 
this part.
    (d) NRC will revise the package identification number to designate 
previously approved package designs as B(U), B(M), AF, BF, or A as 
appropriate, and with the identification number suffix ``-85'' after 
receipt of an application demonstrating that the design meets the 
requirements of this part.


Sec. 71.14  General license: DOT specification container.

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


Sec. 71.16  General License: Use of foreign approved package.

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

[[Page 50269]]
this part. With respect to the quality assurance provisions of subpart 
H of this part, the licensee is exempt from design, construction, and 
fabrication considerations.


Sec. 71.18  General license: Fissile material, limited quantity per 
package.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, without complying with the package standards of 
subparts E and F of this part, if the material is shipped in accordance 
with this section.
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only when a package contains no 
more than a Type A quantity of radioactive material, including only one 
of the following:
    (1) Up to 40 g of uranium-235;
    (2) Up to 30 g of uranium-233;
    (3) Up to 25 g of the fissile radionuclides of plutonium, except 
that for encapsulated plutonium-beryllium neutron sources in special 
form, an A1 quantity of plutonium may be present; or
    (4) A combination of fissile radionuclides in which the sum of the 
ratios of the amount of each radionuclide to the corresponding maximum 
amounts in paragraphs (c)(1), (2), and (3) of this section does not 
exceed unity.
    (d) (1) This general license applies only when, except as specified 
below for encapsulated plutonium-beryllium sources, a package 
containing more than 15 g of fissile radionuclides is labeled with a 
transport index not less than the number given by the following 
equation, where the package contains x grams of uranium-235, y grams of 
uranium-233, and z grams of the fissile radionuclides of plutonium:

Minimum Transport Index = (0.40x+0.67y+z) (1-15 ).x+y+z

    (2) For a package in which the only fissile material is in the form 
of encapsulated plutonium-beryllium neutron sources in special form, 
the transport index based on criticality considerations may be taken as 
0.026 times the number of grams of the fissile radionuclides of 
plutonium in excess of 15 g. In all cases, the transport index must be 
rounded up to one decimal place and may not exceed 10.0.


Sec. 71.20  General license: Fissile material, limited moderator per 
package.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, without complying with the package standards of 
subparts E and F of this part if the material is shipped in accordance 
with this section.
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only when--
    (1) The package contains no more than a Type A quantity of 
radioactive material;
    (2) Neither beryllium nor hydrogenous material enriched in 
deuterium is present;
    (3) The total mass of graphite present does not exceed 7.7 times 
the total mass of uranium-235 plus plutonium;
    (4) Substances having a higher hydrogen density than water (e.g., 
certain hydrocarbon oils), are not present, except that polyethylene 
may be used for packing or wrapping;
    (5) Uranium-233 is not present, and the amount of plutonium does 
not exceed 1 percent of the amount of uranium-235;
    (6) The amount of uranium-235 is limited as follows:
    (i) If the fissile radionuclides are not uniformly distributed, the 
maximum amount of uranium-235 per package may not exceed the value 
given in Table I of this part; or
    (ii) If the fissile radionuclides are distributed uniformly (i.e., 
cannot form a lattice arrangement within the packaging), the maximum 
amount of uranium-235 per package may not exceed the value given in 
Table II of this part; and
    (7) The transport index of each package, based on criticality 
considerations, is taken as 10 times the number of grams of uranium-235 
in the package divided by the maximum allowable number of grams per 
package in accordance with Table I or Table II of this part, as 
applicable.

 Table I.--Permissible Mass of Uranium-235 per Fissile Material Package,
                   Applicable to Sec.  71.20(c)(6)(i)                   
                        [Nonuniform distribution]                       
------------------------------------------------------------------------
                                                             Permissible
                                                               maximum  
  Uranium enrichment in weight percent of uranium-235 not      grams of 
                         exceeding                           uranium-235
                                                             per package
------------------------------------------------------------------------
24.........................................................           40
20.........................................................           42
15.........................................................           45
11.........................................................           48
10.........................................................           51
9.5........................................................           52
9..........................................................           54
8.5........................................................           55
8..........................................................           57
7.5........................................................           59
7..........................................................           60
6.5........................................................           62
6..........................................................           65
5.5........................................................           68
5..........................................................           72
4.5........................................................           76
4..........................................................           80
3.5........................................................           88
3..........................................................          100
2.5........................................................          120
2..........................................................          164
1.5........................................................          272
1.35.......................................................          320
1..........................................................          680
0.92.......................................................        1,200
------------------------------------------------------------------------


Table II.--Permissible Mass of Uranium-235 per Fissile Material Package,
                   Applicable to Sec.  71.20(c)(6)(ii)                  
                         [Uniform Distribution]                         
------------------------------------------------------------------------
                                                             Permissible
                                                               maximum  
  Uranium enrichment in weight percent of uranium-235 not      grams of 
                         exceeding                           uranium-235
                                                             per package
------------------------------------------------------------------------
4..........................................................           84
3.5........................................................           92
3..........................................................          112
2.5........................................................          148
2..........................................................          240
1.5........................................................          560
1.35.......................................................          800
------------------------------------------------------------------------

Sec. 71.22 General license:  Fissile material, limited quantity, 
controlled shipment.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, without complying with the package standards of 
Subparts E and F of this part, if limited material is shipped in 
accordance with this section.
    (b) The general license applies only to a licensee who has a 
quality assurance 

[[Page 50270]]
program approved by the Commission as satisfying the provisions of 
Subpart H of this part.
    (c) This general license applies only when a package contains no 
more than a Type A quantity of radioactive material and no more than 
400 g total of the fissile radionuclides of plutonium encapsulated as 
plutonium-beryllium neutron sources in special form.
    (d) This general license applies only when the fissile 
radionuclides in the shipment exceed none of the following:
    (1) 500 g of uranium-235;
    (2) 300 g total of uranium-233, and the fissile radionuclides of 
plutonium;
    (3) A total quantity of uranium-233, uranium-235, and the fissile 
radionuclides of plutonium so that the sum of the ratios of the 
quantity of each radionuclide to the quantity specified in paragraphs 
(d)(1) and (d)(2) of this section does not exceed unity; or
    (4) 2500 g total of the fissile radionuclides of plutonium 
encapsulated as plutonium-beryllium neutron sources in special form.
    (e) This general license applies only when shipment of these 
packages is made under procedures specifically authorized by DOT, in 
accordance with 49 CFR part 173 of its regulations, to prevent loading, 
transport, or storage of these packages with other fissile material 
shipments.


Sec. 71.24  General license: Fissile material, limited moderator, 
controlled shipment.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, without complying with the package standards of 
subparts E and F of this part, if limited material is shipped in 
accordance with this section.
    (b) The general license applies only to a licensee who has a 
quality assurance program approved by the Commission as satisfying the 
provisions of subpart H of this part.
    (c) This general license applies only when--
    (1) No package contains more than a Type A quantity of radioactive 
material;
    (2) The packaging does not incorporate lead shielding exceeding 5 
cm in thickness, tungsten shielding, or uranium shielding;
    (3) Neither beryllium nor hydrogenous material enriched in 
deuterium is present;
    (4) The total mass of graphite present does not exceed 7.7 times 
the total mass of uranium-235 and plutonium;
    (5) Substances having a higher hydrogen density than water (e.g., 
certain hydrocarbon oils), are not present, except that polyethylene 
may be used for packing or wrapping;
    (6) For fissile contents containing no uranium-233 and less than 1 
percent by weight total plutonium, if the fissile radionuclides are--
    (i) Not uniformly distributed, the maximum amount of uranium-235 
per consignment does not exceed the value given in Table III of this 
part; or
    (ii) Distributed uniformly and cannot form a lattice arrangement 
within the packaging, the maximum amount of uranium-235 per shipment 
does not exceed the value given in Table IV of this part;
    (7) For fissile contents containing uranium-233 or more than 1 
percent by weight plutonium, the total mass of fissile material per 
shipment is limited so that the sum of the number of grams of uranium-
235 divided by 400, the number of grams of plutonium divided by 225, 
and the number of grams of uranium-233 divided by 250, does not exceed 
unity, as expressed in the formula:
[GRAPHIC][TIFF OMITTED]TR28SE95.000

    (8) The transport must be direct to the consignee without any 
intermediate transit storage; and
    (9) Shipment of these packages is made under procedures 
specifically authorized by DOT in accordance with 49 CFR part 173 of 
its regulations to prevent loading, transport, or storage of these 
packages with other fissile material shipments.

    Table III.--Permissible Mass of Uranium-235 per Fissile Material    
               Shipment Applicable to Sec.  71.24(c)(6)(i)              
                        [Nonuniform distribution]                       
------------------------------------------------------------------------
                                                            Permissible 
   Uranium enrichment in weight percent of uranium-235     maximum grams
                      notexceeding                        of uranium-235
                                                          perconsignment
------------------------------------------------------------------------
20......................................................            520 
15......................................................            560 
11......................................................            600 
10......................................................            640 
9.5.....................................................            655 
9.......................................................            675 
8.5.....................................................            690 
8.......................................................            710 
7.5.....................................................            730 
7.......................................................            750 
6.5.....................................................            780 
6.......................................................            810 
5.5.....................................................            850 
5.......................................................            900 
4.5.....................................................            950 
4.......................................................          1,000 
3.5.....................................................          1,100 
3.......................................................          1,250 
2.5.....................................................          1,500 
2.......................................................          2,050 
1.5.....................................................          3,400 
1.35....................................................          4,000 
1.......................................................          8,500 
0.92....................................................         15,000 
------------------------------------------------------------------------


Table IV.--Permissible Mass of Uranium-235 per Fissile Material Shipment
                   Applicable to Sec.  71.24(c)(6)(ii)                  
                         [Uniform distribution]                         
------------------------------------------------------------------------
                                                             Permissible
                                                               maximum  
  Uranium enrichment in weight percent of uranium-235 not     grams of  
                         exceeding                           uranium-235
                                                                 per    
                                                             consignment
------------------------------------------------------------------------
4.........................................................        1,050 
3.5.......................................................        1,150 
3.........................................................        1,400 
2.5.......................................................        1,800 
2.........................................................        3,000 
1.5.......................................................        7,000 
1.35......................................................       10,000 
------------------------------------------------------------------------

Subpart D--Application for Package Approval


Sec. 71.31  Contents of application.

    (a) An application for an approval under this part must include, 
for each proposed packaging design, the following information:
    (1) A package description as required by Sec. 71.33; 

[[Page 50271]]

    (2) A package evaluation as required by Sec. 71.35; and
    (3) A quality assurance program description, as required by 
Sec. 71.37, or a reference to a previously approved quality assurance 
program.
    (b) Except as provided in Sec. 71.13, an application for 
modification of a package design, whether for modification of the 
packaging or authorized contents, must include sufficient information 
to demonstrate that the proposed design satisfies the package standards 
in effect at the time the application is filed.
    (c) The applicant shall identify any established codes and 
standards proposed for use in package design, fabrication, assembly, 
testing, maintenance, and use. In the absence of any codes and 
standards, the applicant shall describe and justify the basis and 
rationale used to formulate the package quality assurance program.


Sec. 71.33  Package description.

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


Sec. 71.35  Package evaluation.

    The application must include the following:
    (a) A demonstration that the package satisfies the standards 
specified in subparts E and F of this part;
    (b) For a fissile material package, the allowable number of 
packages that may be transported in the same vehicle in accordance with 
Sec. 71.59; and
    (c) For a fissile material shipment, any proposed special controls 
and precautions for transport, loading, unloading, and handling and any 
proposed special controls in case of an accident or delay.


Sec. 71.37  Quality assurance.

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


Sec. 71.38  Renewal of a certificate of compliance or quality assurance 
program approval.

    (a) Except as provided in paragraph (b) of this section, each 
Certificate of Compliance or Quality Assurance Program Approval 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 or Quality 
Assurance Program Approval issued pursuant to the part, has filed an 
application in proper form for renewal of either of those approvals, 
the existing Certificate of Compliance or Quality Assurance Program 
Approval 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 or Quality Assurance Program Approval, an applicant may be 
required to submit a consolidated application that incorporates all 
changes to its program that, are incorporated by reference in the 
existing approval or certificate, into as few referenceable documents 
as reasonably achievable.


Sec. 71.39  Requirement for additional information.

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

Subpart E--Package Approval Standards


Sec. 71.41  Demonstration of compliance.

    (a) The effects on a package of the tests specified in Sec. 71.71 
(``Normal conditions of transport''), and the tests specified in 
Sec. 71.73 (``Hypothetical accident conditions''), and Sec. 71.61 
(Special requirement for irradiated nuclear fuel shipments''), must be 
evaluated by subjecting a specimen or scale model to a specific test, 
or by another method of demonstration acceptable to the Commission, as 
appropriate for the particular feature being considered.
    (b) Taking into account the type of vehicle, the method of securing 
or attaching the package, and the controls to be exercised by the 
shipper, the Commission may permit the shipment to be evaluated 
together with the transporting vehicle.
    (c) Environmental and test conditions different from those 
specified in Secs. 71.71 and 71.73 may be approved by the Commission if 
the controls proposed to be exercised by the shipper are demonstrated 
to be adequate to provide equivalent safety of the shipment.


Sec. 71.43  General standards for all packages.

    (a) The smallest overall dimension of a package may not be less 
than 10 cm (4 in).
    (b) The outside of a package must incorporate a feature, such as a 
seal, that is not readily breakable and that, while intact, would be 
evidence that the package has not been opened by unauthorized persons.
    (c) Each package must include a containment system securely closed 
by a positive fastening device that cannot be opened unintentionally or 
by a pressure that may arise within the package.
    (d) A package must be made of materials and construction that 
assure that there will be no significant chemical, galvanic, or other 
reaction among the packaging components, among package contents, or 
between the packaging components and the package contents, including 
possible reaction resulting from inleakage of water, to the maximum 
credible extent. Account 

[[Page 50272]]
must be taken of the behavior of materials under irradiation.
    (e) A package valve or other device, the failure of which would 
allow radioactive contents to escape, must be protected against 
unauthorized operation and, except for a pressure relief device, must 
be provided with an enclosure to retain any leakage.
    (f) A package must be designed, constructed, and prepared for 
shipment so that under the tests specified in Sec. 71.71 (``Normal 
conditions of transport'') there would be no loss or dispersal of 
radioactive contents, no significant increase in external surface 
radiation levels, and no substantial reduction in the effectiveness of 
the packaging.
    (g) A package must be designed, constructed, and prepared for 
transport so that in still air at 38 deg.C (100 deg.F) and in the 
shade, no accessible surface of a package would have a temperature 
exceeding 50 deg.C (122 deg.F) in a nonexclusive use shipment, or 
85 deg.C (185 deg.F) in an exclusive use shipment.
    (h) A package may not incorporate a feature intended to allow 
continuous venting during transport.


Sec. 71.45  Lifting and tie-down standards for all packages.

    (a) Any lifting attachment that is a structural part of a package 
must be designed with a minimum safety factor of three against yielding 
when used to lift the package in the intended manner, and it must be 
designed so that failure of any lifting device under excessive load 
would not impair the ability of the package to meet other requirements 
of this subpart. Any other structural part of the package that could be 
used to lift the package must be capable of being rendered inoperable 
for lifting the package during transport, or must be designed with 
strength equivalent to that required for lifting attachments.
    (b) Tie-down devices:
    (1) If there is a system of tie-down devices that is a structural 
part of the package, the system must be capable of withstanding, 
without generating stress in any material of the package in excess of 
its yield strength, a static force applied to the center of gravity of 
the package having a vertical component of 2 times the weight of the 
package with its contents, a horizontal component along the direction 
in which the vehicle travels of 10 times the weight of the package with 
its contents, and a horizontal component in the transverse direction of 
5 times the weight of the package with its contents.
    (2) Any other structural part of the package that could be used to 
tie down the package must be capable of being rendered inoperable for 
tying down the package during transport, or must be designed with 
strength equivalent to that required for tie-down devices.
    (3) Each tie-down device that is a structural part of a package 
must be designed so that failure of the device under excessive load 
would not impair the ability of the package to meet other requirements 
of this part.


Sec. 71.47  External radiation standards for all packages.

    (a) Except as provided in paragraph (b) of this section, each 
package of radioactive materials offered for transportation must be 
designed and prepared for shipment so that under conditions normally 
incident to transportation the radiation level does not exceed 2 mSv/h 
(200 mrem/h) at any point on the external surface of the package, and 
the transport index does not exceed 10.
    (b) A package that exceeds the radiation level limits specified in 
paragraph (a) of this section must be transported by exclusive use 
shipment only, and the radiation levels for such shipment must not 
exceed the following during transportation:
    (1) 2 mSv/h (200 mrem/h) on the external surface of the package, 
unless the following conditions are met, in which case the limit is 10 
mSv/h (1000 mrem/h):
    (i) The shipment is made in a closed transport vehicle;
    (ii) The package is secured within the vehicle so that its position 
remains fixed during transportation; and
    (iii) There are no loading or unloading operations between the 
beginning and end of the transportation;
    (2) 2 mSv/h (200 mrem/h) at any point on the outer surface of the 
vehicle, including the top and underside of the vehicle; or in the case 
of a flat-bed style vehicle, at any point on the vertical planes 
projected from the outer edges of the vehicle, on the upper surface of 
the load or enclosure, if used, and on the lower external surface of 
the vehicle; and
    (3) 0.1 mSv/h (10 mrem/h) at any point 2 meters (80 in) from the 
outer lateral surfaces of the vehicle (excluding the top and underside 
of the vehicle); or in the case of a flat-bed style vehicle, at any 
point 2 meters (6.6 feet) from the vertical planes projected by the 
outer edges of the vehicle (excluding the top and underside of the 
vehicle); and
    (4) 0.02 mSv/h (2 mrem/h) in any normally occupied space, except 
that this provision does not apply to private carriers, if exposed 
personnel under their control wear radiation dosimetry devices in 
conformance with 10 CFR 20.1502.
    (c) For shipments made under the provisions of paragraph (b) of 
this section, the shipper shall provide specific written instructions 
to the carrier for maintenance of the exclusive use shipment controls. 
The instructions must be included with the shipping paper information.
    (d) The written instructions required for exclusive use shipments 
must be sufficient so that, when followed, they will cause the carrier 
to avoid actions that will unnecessarily delay delivery or 
unnecessarily result in increased radiation levels or radiation 
exposures to transport workers or members of the general public.


Sec. 71.51  Additional requirements for Type B packages.

    (a) Except as provided in Sec. 71.52, a Type B package, in addition 
to satisfying the requirements of Secs. 71.41 through 71.47, must be 
designed, constructed, and prepared for shipment so that under the 
tests specified in:
    (1) Section 71.71 (``Normal conditions of transport''), there would 
be no loss or dispersal of radioactive contents--as demonstrated to a 
sensitivity of 10-6 A2 per hour, no significant increase in 
external surface radiation levels, and no substantial reduction in the 
effectiveness of the packaging; and
    (2) Section 71.73 (``Hypothetical accident conditions''), there 
would be no escape of krypton-85 exceeding 10 A2 in 1 week, no 
escape of other radioactive material exceeding a total amount A2 
in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 
rem/h) at 1 m (40 in) from the external surface of the package.
    (b) Where mixtures of different radionuclides are present, the 
provisions of appendix A, paragraph IV of this part shall apply, except 
that for Krypton-85, an effective A2 value equal to 10 A2 may 
be used.
    (c) Compliance with the permitted activity release limits of 
paragraph (a) of this section may not depend on filters or on a 
mechanical cooling system.


Sec. 71.52  Exemption for low-specific-activity (LSA) packages.

    A package need not satisfy the requirements of Sec. 71.51 if it 
contains only LSA or SCO material, and is transported as exclusive use, 
but is subject to Secs. 71.41 through 71.47, including Sec. 71.43(f). 
This section expires April 1, 1999.


Sec. 71.53  Fissile material exemptions.

    The following packages are exempt from fissile material 
classification and 

[[Page 50273]]
from the fissile material standards of Sec. 71.55 and Sec. 71.59, but 
are subject to all other requirements of this part:
    (a) A package containing not more than 15 g of fissile material. If 
material is transported in bulk, the quantity limitation applies to the 
conveyance;
    (b) A package containing homogeneous hydrogenous solutions or 
mixtures where:
    (1) The minimum ratio of the number of hydrogen atoms to the number 
of atoms of fissile radionuclides (H/X) is 5200;
    (2) The maximum concentration of fissile radionuclides is 5 g/
liter; and
    (3) The maximum mass of fissile radionuclides in the package is 800 
g, with an exception for a mixture where the total mass of plutonium 
and uranium-233 exceeds 1 percent of the mass of uranium-235, the limit 
is 500 g. If the material is transported in bulk, other than by 
aircraft, the quantity limitations apply to the conveyance;
    (c) A package containing uranium enriched in uranium-235 to a 
maximum of 1 percent by weight, and with a total plutonium and uranium-
233 content of up to 1 percent of the mass of uranium-235, if the 
fissile radionuclides are distributed homogeneously throughout the 
package contents and do not form a lattice arrangement within the 
package;
    (d) A package containing any fissile material if it does not 
contain more than 5 g of fissile radionuclides in any 10 liter volume, 
and if the material is packaged so as to maintain this limit of fissile 
radionuclide concentration during normal transport;
    (e) A package containing not more than 1 kg of plutonium of which 
not more than 20 percent by mass may consist of plutonium-239, 
plutonium-241, or any combination of those radionuclides; or
    (f) A package containing liquid solutions of uranyl nitrate 
enriched in uranium-235 to a maximum of 2 percent by weight, with total 
plutonium and uranium-233 not more than 0.1 percent of the mass of 
uranium-235 and with a minimum nitrogen-to-uranium atomic ratio (N/U) 
of 2.


Sec. 71.55  General requirements for fissile material packages.

    (a) A package used for the shipment of fissile material must be 
designed and constructed in accordance with Secs. 71.41 through 71.47. 
When required by the total amount of radioactive material, a package 
used for the shipment of fissile material must also be designed and 
constructed in accordance with Sec. 71.51.
    (b) Except as provided in paragraph (c) of this section, a package 
used for the shipment of fissile material must be so designed and 
constructed and its contents so limited that it would be subcritical if 
water were to leak into the containment system, or liquid contents were 
to leak out of the containment system so that, under the following 
conditions, maximum reactivity of the fissile material would be 
attained:
    (1) The most reactive credible configuration consistent with the 
chemical and physical form of the material;
    (2) Moderation by water to the most reactive credible extent; and
    (3) Close full reflection of the containment system by water on all 
sides, or such greater reflection of the containment system as may 
additionally be provided by the surrounding material of the packaging.
    (c) The Commission may approve exceptions to the requirements of 
paragraph (b) of this section if the package incorporates special 
design features that ensure that no single packaging error would permit 
leakage, and if appropriate measures are taken before each shipment to 
ensure that the containment system does not leak.
    (d) A package used for the shipment of fissile material must be so 
designed and constructed and its contents so limited that under the 
tests specified in Sec. 71.71 (``Normal conditions of transport'')--
    (1) The contents would be subcritical;
    (2) The geometric form of the package contents would not be 
substantially altered;
    (3) There would be no leakage of water into the containment system 
unless, in the evaluation of undamaged packages under Sec. 71.59(b)(1), 
it has been assumed that moderation is present to such an extent as to 
cause maximum reactivity consistent with the chemical and physical form 
of the material; and
    (4) There will be no substantial reduction in the effectiveness of 
the packaging, including:
    (i) No more than 5 percent reduction in the total effective volume 
of the packaging on which nuclear safety is assessed;
    (ii) No more than 5 percent reduction in the effective spacing 
between the fissile contents and the outer surface of the packaging; 
and
    (iii) No occurrence of an aperture in the outer surface of the 
packaging large enough to permit the entry of a 10 cm (4 in) cube.
    (e) A package used for the shipment of fissile material must be so 
designed and constructed and its contents so limited that under the 
tests specified in Sec. 71.73 (``Hypothetical accident conditions''), 
the package would be subcritical. For this determination, it must be 
assumed that:
    (1) The fissile material is in the most reactive credible 
configuration consistent with the damaged condition of the package and 
the chemical and physical form of the contents;
    (2) Water moderation occurs to the most reactive credible extent 
consistent with the damaged condition of the package and the chemical 
and physical form of the contents; and
    (3) There is full reflection by water on all sides, as close as is 
consistent with the damaged condition of the package.


Sec. 71.57  [Reserved]


Sec. 71.59  Standards for arrays of fissile material packages.

    (a) A fissile material package must be controlled by either the 
shipper or the carrier during transport to assure that an array of such 
packages remains subcritical. To enable this control, the designer of a 
fissile material package shall derive a number ``N'' based on all the 
following conditions being satisfied, assuming packages are stacked 
together in any arrangement and with close full reflection on all sides 
of the stack by water:
    (1) Five times ``N'' undamaged packages with nothing between the 
packages would be subcritical;
    (2) Two times ``N'' damaged packages, if each package were 
subjected to the tests specified in Sec. 71.73 (``Hypothetical accident 
conditions'') would be subcritical with optimum interspersed 
hydrogenous moderation; and
    (3) The value of ``N'' cannot be less than 0.5.
     (b) The transport index based on nuclear criticality control must 
be obtained by dividing the number 50 by the value of ``N'' derived 
using the procedures specified in paragraph (a) of this section. The 
value of the transport index for nuclear criticality control may be 
zero provided that an unlimited number of packages is subcritical such 
that the value of ``N'' is effectively equal to infinity under the 
procedures specified in paragraph (a) of this section. Any transport 
index greater than zero must be rounded up to the first decimal place.
     (c) Where a fissile material package is assigned a nuclear 
criticality control transport index--
    (1) Not in excess of 10, that package may be shipped by any 
carrier, and that carrier provides adequate criticality control by 
limiting the sum of the transport indexes to 50 in a non-exclusive use 
vehicle, and to 100 in an exclusive use vehicle.
    (2) In excess of 10, that package may only be shipped by exclusive 
use 

[[Page 50274]]
vehicle or other shipper controlled system specified by DOT for fissile 
material packages. The shipper provides adequate criticality control by 
limiting the sum of the transport indexes to 100 in an exclusive use 
vehicle.


Sec. 71.61  Special requirement for irradiated nuclear fuel shipments.

    A package for irradiated nuclear fuel with activity greater than 37 
PBq (106 Ci) must be so designed that its undamaged containment 
system can withstand an external water pressure of 2 MPa (290 psi) for 
a period of not less than one hour without collapse, buckling, or 
inleakage of water.


Sec. 71.63  Special requirements for plutonium shipments.

    (a) Plutonium in excess of 20 Ci (0.74 TBq) per package must be 
shipped as a solid.
    (b) Plutonium in excess of 20 Ci (0.74 TBq) per package must be 
packaged in a separate inner container placed within outer packaging 
that meets the requirements of subparts E and F of this part for 
packaging of material in normal form. If the entire package is 
subjected to the tests specified in Sec. 71.71 (``Normal conditions of 
transport''), the separate inner container must not release plutonium 
as demonstrated to a sensitivity of 10-6 A2/h. If the entire 
package is subjected to the tests specified in Sec. 71.73 
(``Hypothetical accident conditions''), the separate inner container 
must restrict the loss of plutonium to not more than A2 in 1 week. 
Solid plutonium in the following forms is exempt from the requirements 
of this paragraph:
    (1) Reactor fuel elements;
    (2) Metal or metal alloy; and
    (3) Other plutonium bearing solids that the Commission determines 
should be exempt from the requirements of this section.


Sec. 71.64  Special requirements for plutonium air shipments.

    (a) A package for the shipment of plutonium by air subject to 
Sec. 71.88(a)(4), in addition to satisfying the requirements of 
Secs. 71.41 through 71.63, as applicable, must be designed, 
constructed, and prepared for shipment so that under the tests 
specified in--
    (1) Section 71.74 (``Accident conditions for air transport of 
plutonium'')--
    (i) The containment vessel would not be ruptured in its post-tested 
condition, and the package must provide a sufficient degree of 
containment to restrict accumulated loss of plutonium contents to not 
more than an A2 quantity in a period of 1 week;
    (ii) The external radiation level would not exceed 10 mSv/h (1 rem/
h) at a distance of 1 m (40 in) from the surface of the package in its 
post-tested condition in air; and
    (iii) A single package and an array of packages are demonstrated to 
be subcritical in accordance with this part, except that the damaged 
condition of the package must be considered to be that which results 
from the plutonium accident tests in Sec. 71.74, rather than the 
hypothetical accident tests in Sec. 71.73; and
    (2) Section 71.74(c), there would be no detectable leakage of water 
into the containment vessel of the package.
    (b) With respect to the package requirements of paragraph (a), 
there must be a demonstration or analytical assessment showing that--
    (1) The results of the physical testing for package qualification 
would not be adversely affected to a significant extent by--
    (i) The presence, during the tests, of the actual contents that 
will be transported in the package; and
    (ii) Ambient water temperatures ranging from 0.6 deg.C (+33 deg.F) 
to 38 deg.C (+100 deg.F) for those qualification tests involving water, 
and ambient atmospheric temperatures ranging from -40 deg.C (-40 deg.F) 
to +54 deg.C (+130 deg.F) for the other qualification tests.
    (2) The ability of the package to meet the acceptance standards 
prescribed for the accident condition sequential tests would not be 
adversely affected if one or more tests in the sequence were deleted.


Sec. 71.65  Additional requirements.

    The Commission may, by rule, regulation, or order, impose 
requirements on any licensee, in addition to those established in this 
part, as it deems necessary or appropriate to protect public health or 
to minimize danger to life or property.

Subpart F--Package, Special Form, and LSA-III Tests \2\

    \2\ The package standards related to the tests in this subpart 
are contained in subpart E of this part.
---------------------------------------------------------------------------


Sec. 71.71  Normal conditions of transport.

    (a) Evaluation. Evaluation of each package design under normal 
conditions of transport must include a determination of the effect on 
that design of the conditions and tests specified in this section. 
Separate specimens may be used for the free drop test, the compression 
test, and the penetration test, if each specimen is subjected to the 
water spray test before being subjected to any of the other tests.
    (b) Initial conditions. With respect to the initial conditions for 
the tests in this section, the demonstration of compliance with the 
requirements of this part must be based on the ambient temperature 
preceding and following the tests remaining constant at that value 
between -29 deg.C (-20 deg.F) and +38 deg.C (+100 deg.F) which is most 
unfavorable for the feature under consideration. The initial internal 
pressure within the containment system must be considered to be the 
maximum normal operating pressure, unless a lower internal pressure 
consistent with the ambient temperature considered to precede and 
follow the tests is more unfavorable.
    (c) Conditions and tests.
    (1) Heat. An ambient temperature of 38 deg.C (100 deg.F) in still 
air, and insolation according to the following table:

                             Insolation Data                            
------------------------------------------------------------------------
                                                              Total     
                                                          insolation for
              Form and location of surface                  a 12-hour   
                                                          period(g cal/ 
                                                               cm2      
------------------------------------------------------------------------
 Flat surfaces transported horizontally:                                
    Base...............................................  None           
    Other surfaces.....................................  800            
Flat surfaces not transported horizontally.............  200            
Curved surfaces........................................  400            
------------------------------------------------------------------------

    (2) Cold. An ambient temperature of -40 deg.C (-40 deg.F) in still 
air and shade.
    (3) Reduced external pressure. An external pressure of 25 kPa (3.5 
lbf/in2) absolute.
    (4) Increased external pressure. An external pressure of 140 kPa 
(20 lbf/in2) absolute.
    (5) Vibration. Vibration normally incident to transport.
    (6) Water spray. A water spray that simulates exposure to rainfall 
of approximately 5 cm/h (2 in/h) for at least 1 hour.
    (7) Free drop. Between 1.5 and 2.5 hours after the conclusion of 
the water spray test, a free drop through the distance specified below 
onto a flat, essentially unyielding, horizontal surface, striking the 
surface in a position for which maximum damage is expected.

              Criteria for Free Drop Test (Weight/Distance)             
------------------------------------------------------------------------
                    Package weight                          Free drop   
-------------------------------------------------------     distance    
                                                       -----------------
         Kilograms                   (Pounds)            Meters   (Feet)
------------------------------------------------------------------------
Less than 5,000...........  (Less than 11,000)........      1.2     (4) 

[[Page 50275]]
                                                                        
5,000 to 10,000...........  (11,000 to 22,000)........      0.9      (3)
10,000 to 15,000..........  (22,000 to 33,100)........      0.6      (2)
More than 15,000..........  (More than 33,100)........      0.3      (1)
------------------------------------------------------------------------


    (8) Corner drop. A free drop onto each corner of the package in 
succession, or in the case of a cylindrical package onto each quarter 
of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially 
unyielding, horizontal surface. This test applies only to fiberboard, 
wood, or fissile material rectangular packages not exceeding 50 kg (110 
lbs) and fiberboard, wood, or fissile material cylindrical packages not 
exceeding 100 kg (220 lbs).
    (9) Compression. For packages weighing up to 5000 kg (11,000 lbs), 
the package must be subjected, for a period of 24 hours, to a 
compressive load applied uniformly to the top and bottom of the package 
in the position in which the package would normally be transported. The 
compressive load must be the greater of the following:
    (i) The equivalent of 5 times the weight of the package; or
    (ii) The equivalent of 13 kPa (2 lbf/in2) multiplied by the 
vertically projected area of the package.
    (10) Penetration. Impact of the hemispherical end of a vertical 
steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, 
dropped from a height of 1 m (40 in) onto the exposed surface of the 
package that is expected to be most vulnerable to puncture. The long 
axis of the cylinder must be perpendicular to the package surface.


Sec. 71.73  Hypothetical accident conditions.

    (a) Test procedures. Evaluation for hypothetical accident 
conditions is to be based on sequential application of the tests 
specified in this section, in the order indicated, to determine their 
cumulative effect on a package or array of packages. An undamaged 
specimen may be used for the water immersion tests specified in 
paragraph (c)(6) of this section.
    (b) Test conditions. With respect to the initial conditions for the 
tests, except for the water immersion tests, to demonstrate compliance 
with the requirements of this part during testing, the ambient air 
temperature before and after the tests must remain constant at that 
value between -29 deg.C (-20 deg.F) and +38 deg.C (+100 deg.F) which is 
most unfavorable for the feature under consideration. The initial 
internal pressure within the containment system must be the maximum 
normal operating pressure, unless a lower internal pressure, consistent 
with the ambient temperature assumed to precede and follow the tests, 
is more unfavorable.
    (c) Tests. Tests for hypothetical accident conditions must be 
conducted as follows:
    (1) Free Drop. A free drop of the specimen through a distance of 9 
m (30 ft) onto a flat, essentially unyielding, horizontal surface, 
striking the surface in a position for which maximum damage is 
expected.
    (2) Crush. Subjection of the specimen to a dynamic crush test by 
positioning the specimen on a flat, essentially unyielding, horizontal 
surface so as to suffer maximum damage by the drop of a 500 kg (1100 
pound) mass from 9 m (30 ft) onto the specimen. The mass must consist 
of a solid mild steel plate 1 m (40 in) by 1 m and must fall in a 
horizontal attitude. The crush test is required only when the specimen 
has a mass not greater than 500 kg (1100 lbs), an overall density not 
greater than 1000 kg/m3 (62.4 lbs/ft3) based on external 
dimensions, and radioactive contents greater than 1000 A2 not as 
special form radioactive material.
    (3) Puncture. A free drop of the specimen through a distance of 1 m 
(40 in) in a position for which maximum damage is expected, onto the 
upper end of a solid, vertical, cylindrical, mild steel bar mounted on 
an essentially unyielding, horizontal surface. The bar must be 15 cm (6 
in) in diameter, with the top horizontal and its edge rounded to a 
radius of not more than 6 mm (0.25 in), and of a length as to cause 
maximum damage to the package, but not less than 20 cm (8 in) long. The 
long axis of the bar must be vertical.
    (4) Thermal. Exposure of the specimen fully engulfed, except for a 
simple support system, in a hydrocarbon fuel/air fire of sufficient 
extent, and in sufficiently quiescent ambient conditions, to provide an 
average emissivity coefficient of at least 0.9, with an average flame 
temperature of at least 800 deg.C (1475 deg.F) for a period of 30 
minutes, or any other thermal test that provides the equivalent total 
heat input to the package and which provides a time averaged 
environmental temperature of 800 deg.C. The fuel source must extend 
horizontally at least 1 m (40 in), but may not extend more than 3 m (10 
ft), beyond any external surface of the specimen, and the specimen must 
be positioned 1 m (40 in) above the surface of the fuel source. For 
purposes of calculation, the surface absorptivity coefficient must be 
either that value which the package may be expected to possess if 
exposed to the fire specified or 0.8, whichever is greater; and the 
convective coefficient must be that value which may be demonstrated to 
exist if the package were exposed to the fire specified. Artificial 
cooling may not be applied after cessation of external heat input, and 
any combustion of materials of construction, must be allowed to proceed 
until it terminates naturally.
    (5) Immersion--fissile material. For fissile material subject to 
Sec. 71.55, in those cases where water inleakage has not been assumed 
for criticality analysis, immersion under a head of water of at least 
0.9 m (3 ft) in the attitude for which maximum leakage is expected.
    (6) Immersion--all packages. A separate, undamaged specimen must be 
subjected to water pressure equivalent to immersion under a head of 
water of at least 15 m (50 ft). For test purposes, an external pressure 
of water of 150 kPa (21.7 lbf/in2) gauge is considered to meet 
these conditions.


Sec. 71.74  Accident conditions for air transport of plutonium.

     (a) Test conditions--Sequence of tests. A package must be 
physically tested to the following conditions in the order indicated to 
determine their cumulative effect.
    (1) Impact at a velocity of not less than 129 m/sec (422 ft/sec) at 
a right angle onto a flat, essentially unyielding, horizontal surface, 
in the orientation (e.g., side, end, corner) expected to result in 
maximum damage at the conclusion of the test sequence.
    (2) A static compressive load of 31,800 kg (70,000 lbs) applied in 
the orientation expected to result in maximum damage at the conclusion 
of the test sequence. The force on the package must be developed 
between a flat steel surface and a 5 cm (2 in) wide, straight, solid, 
steel bar. The length of the bar must be at least as long as the 
diameter of the package, and the longitudinal axis of the bar must be 
parallel to the plane of the flat surface. The load must be applied to 
the bar in a manner that prevents any members or devices used to 
support the bar from contacting the package.
    (3) Packages weighing less than 227 kg (500 lbs) must be placed on 
a flat, essentially unyielding, horizontal surface, and subjected to a 
weight of 227 kg (500 lbs) falling from a height of 3 m (10 ft) and 
striking in the position expected to result in maximum damage at the 
conclusion of the test sequence. 

[[Page 50276]]
The end of the weight contacting the package must be a solid probe made 
of mild steel. The probe must be the shape of the frustum of a right 
circular cone, 30 cm (12 in) long, 20 cm (8 in) in diameter at the 
base, and 2.5 cm (1 in) in diameter at the end. The longitudinal axis 
of the probe must be perpendicular to the horizontal surface. For 
packages weighing 227 kg (500 lbs) or more, the base of the probe must 
be placed on a flat, essentially unyielding horizontal surface, and the 
package dropped from a height of 3 m (10 ft) onto the probe, striking 
in the position expected to result in maximum damage at the conclusion 
of the test sequence.
    (4) The package must be firmly restrained and supported such that 
its longitudinal axis is inclined approximately 45 deg. to the 
horizontal. The area of the package that made first contact with the 
impact surface in paragraph (a)(1) of this section must be in the 
lowermost position. The package must be struck at approximately the 
center of its vertical projection by the end of a structural steel 
angle section falling from a height of at least 46 m (150 ft). The 
angle section must be at least 1.8 m (6 ft) in length with equal legs 
at least 13 cm (5 in) long and 1.3 cm (0.5 in) thick. The angle section 
must be guided in such a way as to fall end-on, without tumbling. The 
package must be rotated approximately 90 deg. about its longitudinal 
axis and struck by the steel angle section falling as before.
    (5) The package must be exposed to luminous flames from a pool fire 
of JP-4 or JP-5 aviation fuel for a period of at least 60 minutes. The 
luminous flames must extend an average of at least 0.9 m (3 ft) and no 
more than 3 m (10 ft) beyond the package in all horizontal directions. 
The position and orientation of the package in relation to the fuel 
must be that which is expected to result in maximum damage at the 
conclusion of the test sequence. An alternate method of thermal testing 
may be substituted for this fire test, provided that the alternate test 
is not of shorter duration and would not result in a lower heating rate 
to the package. At the conclusion of the thermal test, the package must 
be allowed to cool naturally or must be cooled by water sprinkling, 
whichever is expected to result in maximum damage at the conclusion of 
the test sequence.
    (6) Immersion under at least 0.9 m (3 ft) of water.
    (b) Individual free-fall impact test.
    (1) An undamaged package must be physically subjected to an impact 
at a velocity not less than the calculated terminal free-fall velocity, 
at mean sea level, at a right angle onto a flat, essentially 
unyielding, horizontal surface, in the orientation (e.g., side, end, 
corner) expected to result in maximum damage.
    (2) This test is not required if the calculated terminal free-fall 
velocity of the package is less than 129 m/sec (422 ft/sec), or if a 
velocity not less than either 129 m/sec (422 ft/sec) or the calculated 
terminal free-fall velocity of the package is used in the sequential 
test of paragraph (a)(1) of this section.
    (c) Individual deep submersion test. An undamaged package must be 
physically submerged and physically subjected to an external water 
pressure of at least 4 MPa (600 lbs/in \2\).


Sec. 71.75  Qualification of special form radioactive material.

    (a) Special form radioactive materials must meet the test 
requirements of paragraph (b) of this section. Each solid radioactive 
material or capsule specimen to be tested must be manufactured or 
fabricated so that it is representative of the actual solid material or 
capsule that will be transported, with the proposed radioactive content 
duplicated as closely as practicable. Any differences between the 
material to be transported and the test material, such as the use of 
non-radioactive contents, must be taken into account in determining 
whether the test requirements have been met. In addition:
    (1) A different specimen may be used for each of the tests;
    (2) The specimen may not break or shatter when subjected to the 
impact, percussion, or bending tests;
    (3) The specimen may not melt or disperse when subjected to the 
heat test;
    (4) After each test, leaktightness or indispersibility of the 
specimen must be determined by a method no less sensitive than the 
leaching assessment procedure prescribed in paragraph (c) of this 
section. For a capsule resistant to corrosion by water, and which has 
an internal void volume greater than 0.1 milliliter, an alternative to 
the leaching assessment is a demonstration of leaktightness of 
x 10-4 torr-liter/s (1.3 x  x 10-4 atm-cm\3\/s) based on air 
at 25 deg.C (77 deg.F) and one atmosphere differential pressure for 
solid radioactive content, or  x 10-6 torr-liter/s 
(1.3 x  x 10-6 atm-cm\3\/s) for liquid or gaseous radioactive 
content; and
    (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/TR4826-
1979(E), ``Sealed radioactive sources leak test methods'' which is 
available from the American National Standards Institute, 1430 
Broadway, New York, N.Y. 10018.
    (b) Test methods.
    (1) Impact Test. The specimen must fall onto the target from a 
height of 9 m (30 ft) or greater in the orientation expected to result 
in maximum damage. The target must be a flat, horizontal surface of 
such mass and rigidity that any increase in its resistance to 
displacement or deformation, on impact by the specimen, would not 
significantly increase the damage to the specimen.
    (2) Percussion Test.
    (i) The specimen must be placed on a sheet of lead that is 
supported by a smooth solid surface, and struck by the flat face of a 
steel billet so as to produce an impact equivalent to that resulting 
from a free drop of 1.4 kg (3 lbs) through 1 m (40 in);
    (ii) The flat face of the billet must be 25 millimeters (mm) (1 
inch) in diameter with the edges 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 thickness 25 mm (1 in) or greater, and must cover an area 
greater than that covered by the specimen;
    (iv) A fresh surface of lead must be used for each impact; and
    (v) The billet must strike the specimen so as to cause maximum 
damage.
    (3) Bending test.
    (i) This test applies only to long, slender sources with a length 
of 10 cm (4 inches) or greater and a length to width ratio of 10 or 
greater;
    (ii) The specimen must be rigidly clamped in a horizontal position 
so that one half of its length protrudes from the face of the clamp;
    (iii) The orientation of the specimen must be such that the 
specimen will suffer maximum damage when its free end is struck by the 
flat face of a steel billet;
    (iv) The billet must strike the specimen so as to produce an impact 
equivalent to that resulting from a free vertical drop of 1.4 kg (3 
lbs) through 1 m (40 in); and
    (v) The flat face of the billet must be 25 mm (1 inch) in diameter 
with the edges rounded off to a radius of 3 mm  0.3 mm (.12 
in  0.012 in).
    (4) Heat test. The specimen must be heated in air to a temperature 
of not less than 800 deg.C (1475 deg.F), held at that temperature for a 
period of 10 minutes, and then allowed to cool.
    (c) Leaching assessment methods. (1) For indispersible solid 
material--
    (i) The specimen must be immersed for 7 days in water at ambient 

[[Page 50277]]
    temperature. The water must have a pH of 6-8 and a maximum conductivity 
of 10 micromho per centimeter at 20 deg. (68 deg.F);
    (ii) The water with specimen must then be heated to a temperature 
of 50 deg.C  5 deg.C (122 deg.F  9 deg.F) and 
maintained at this temperature for 4 hours.
    (iii) The activity of the water must then be determined;
    (iv) The specimen must then be stored for at least 7 days in still 
air of relative humidity not less than 90 percent at 30 deg.C 
(86 deg.F);
    (v) The specimen must then be immersed in water under the same 
conditions as in paragraph (c)(1)(i) of this section, and the water 
with specimen must be heated to 50 deg.C  5 deg.C 
(122 deg.F  9 deg.F) and maintained at that temperature for 
4 hours;
    (vi) The activity of the water must then be determined. The sum of 
the activities determined here and in paragraph (c)(1)(iii) of this 
section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie 
(Ci)).
    (2) For encapsulated material--
    (i) The specimen must be immersed in water at ambient temperature. 
The water must have a pH of 6-8 and a maximum conductivity of 10 
micromho per centimeter;
    (ii) The water and specimen must be heated to a temperature of 
50 deg.C  5 deg.C (122 deg.F  9 deg.F) and 
maintained at this temperature for 4 hours;
    (iii) The activity of the water must then be determined;
    (iv) The specimen must then be stored for at least 7 days in still 
air at a temperature of 30 deg.C (86 deg.F) or greater;
    (v) The process in paragraph (c)(2)(i), (ii), and (iii) of this 
section must be repeated; and
    (vi) The activity of the water must then be determined. The sum of 
the activities determined here and in paragraph (c)(2)(iii) of this 
section must not exceed 2 kilobecquerels (kBq) (0.05 microcurie 
(Ci)).
    (d) A specimen that comprises or simulates radioactive material 
contained in a sealed capsule need not be subjected to--
    (1) The impact test and the percussion test of this section, 
provided that the specimen is alternatively subjected to the Class 4 
impact test prescribed in ISO 2919-1980(e), ``Sealed Radioactive 
Sources Classification'' (see Sec. 71.75(a)(5) for statement of 
availability); and
    (2) The heat test of this section, provided the specimen is 
alternatively subjected to the Class 6 temperature test specified in 
the International Organization for Standardization document ISO 2919-
1980(e), ``Sealed Radioactive Sources Classification.''


Sec. 71.77  Qualification of LSA-III Material

    (a) LSA-III material must meet the test requirements of paragraph 
(b) of this section. Any differences between the specimen to be tested 
and the material to be transported must be taken into account in 
determining whether the test requirements have been met.
    (b) Leaching Test. (1) The specimen, representing no less than the 
entire contents of the package, must be immersed for 7 days in water at 
ambient temperature;
    (2) The volume of water to be used in the test must be sufficient 
to ensure that at the end of the test period the free volume of the 
unabsorbed and unreacted water remaining will be at least 10% of the 
volume of the specimen itself;
    (3) The water must have an initial pH of 6-8 and a maximum 
conductivity 10 micromho/cm at 20 deg.C (68 deg.F); and
    (4) The total activity of the free volume of water must be measured 
following the 7 day immersion test and must not exceed 0.1 A2.

Subpart G--Operating Controls and Procedures


Sec. 71.81  Applicability of operating controls and procedures.

    A licensee subject to this part, who, under a general or specific 
license, transports licensed material or delivers licensed material to 
a carrier for transport, shall comply with the requirements of this 
subpart G, with the quality assurance requirements of subpart H of this 
part, and with the general provisions of subpart A of this part.


Sec. 71.83  Assumptions as to unknown properties.

    When the isotopic abundance, mass, concentration, degree of 
irradiation, degree of moderation, or other pertinent property of 
fissile material in any package is not known, the licensee shall 
package the fissile material as if the unknown properties have credible 
values that will cause the maximum neutron multiplication.


Sec. 71.85  Preliminary determinations.

    Before the first use of any packaging for the shipment of licensed 
material--
    (a) The licensee 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 licensee 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; and
    (c) The licensee shall conspicuously and durably mark the packaging 
with its model number, serial number, gross weight, and a package 
identification number assigned by NRC. Before applying the model 
number, the licensee shall determine that the packaging has been 
fabricated in accordance with the design approved by the Commission.


Sec. 71.87  Routine determinations.

    Before each shipment of licensed material, the licensee shall 
ensure that the package with its contents satisfies the applicable 
requirements of this part and of the license. The licensee shall 
determine that--
    (a) The package is proper for the contents to be shipped;
    (b) The package is in unimpaired physical condition except for 
superficial defects such as marks or dents;
    (c) Each closure device of the packaging, including any required 
gasket, is properly installed and secured and free of defects;
    (d) Any system for containing liquid is adequately sealed and has 
adequate space or other specified provision for expansion of the 
liquid;
    (e) Any pressure relief device is operable and set in accordance 
with written procedures;
    (f) The package has been loaded and closed in accordance with 
written procedures;
    (g) For fissile material, any moderator or neutron absorber, if 
required, is present and in proper condition;
    (h) Any structural part of the package that could be used to lift 
or tie down the package during transport is rendered inoperable for 
that purpose, unless it satisfies the design requirements of 
Sec. 71.45;
    (i) The level of non-fixed (removable) radioactive contamination on 
the external surfaces of each package offered for shipment is as low as 
reasonably achievable, and within the limits specified in DOT 
regulations in 49 CFR 173.443;
    (j) External radiation levels around the package and around the 
vehicle, if applicable, will not exceed the limits specified in 
Sec. 71.47 at any time during transportation; and
    (k) Accessible package surface temperatures will not exceed the 
limits specified in Sec. 71.43(g) at any time during transportation.


Sec. 71.88  Air transport of plutonium.

    (a) Notwithstanding the provisions of any general licenses and 

[[Page 50278]]
    notwithstanding any exemptions stated directly in this part or included 
indirectly by citation of 49 CFR Chapter I, as may be applicable, the 
licensee shall assure that plutonium in any form, whether for import, 
export, or domestic shipment, is not transported by air or delivered to 
a carrier for air transport unless:
    (1) The plutonium is contained in a medical device designed for 
individual human application; or
    (2) The plutonium is contained in a material in which the specific 
activity is not greater than 0.002 Ci/g (70 Bq/g) of material 
and in which the radioactivity is essentially uniformly distributed; or
    (3) The plutonium is shipped in a single package containing no more 
than an A2 quantity of plutonium in any isotope or form, and is 
shipped in accordance with Sec. 71.5; or
    (4) The plutonium is shipped in a package specifically authorized 
for the shipment of plutonium by air in the Certificate of Compliance 
for that package issued by the Commission.
    (b) Nothing in paragraph (a) of this section is to be interpreted 
as removing or diminishing the requirements of Sec. 73.24 of this 
chapter.
    (c) For a shipment of plutonium by air which is subject to 
paragraph (a)(4) of this section, the licensee shall, through special 
arrangement with the carrier, require compliance with 49 CFR 175.704, 
U.S. Department of Transportation regulations applicable to the air 
transport of plutonium.


Sec. 71.89  Opening instructions.

    Before delivery of a package to a carrier for transport, the 
licensee shall ensure that any special instructions needed to safely 
open the package have been sent to, or otherwise made available to, the 
consignee for the consignee's use in accordance with 10 CFR 20.1906(e).


Sec. 71.91  Records.

    (a) Each licensee shall maintain, for a period of 3 years after 
shipment, a record of each shipment of licensed material not exempt 
under Sec. 71.10, showing where applicable--
    (1) Identification of the packaging by model number and serial 
number;
    (2) Verification that there are no significant defects in the 
packaging, as shipped;
    (3) Volume and identification of coolant;
    (4) Type and quantity of licensed material in each package, and the 
total quantity of each shipment;
    (5) For each item of irradiated fissile material--
    (i) Identification by model number and serial number;
    (ii) Irradiation and decay history to the extent appropriate to 
demonstrate that its nuclear and thermal characteristics comply with 
license conditions; and
    (iii) Any abnormal or unusual condition relevant to radiation 
safety;
    (6) Date of the shipment;
    (7) For fissile packages and for Type B packages, any special 
controls exercised;
    (8) Name and address of the transferee;
    (9) Address to which the shipment was made; and
    (10) Results of the determinations required by Sec. 71.87 and by 
the conditions of the package approval.
    (b) The licensee shall make available to the Commission for 
inspection, upon reasonable notice, all records required by this part. 
Records are only valid if stamped, initialed, or signed and dated by 
authorized personnel or otherwise authenticated.
    (c) The licensee shall maintain sufficient written records to 
furnish evidence of the quality of packaging. The records to be 
maintained include results of the determinations required by 
Sec. 71.85; design, fabrication, and assembly records, results of 
reviews, inspections, tests, and audits; results of monitoring work 
performance and materials analyses; and results of maintenance, 
modification and repair activities. Inspection, test, and audit records 
must identify the inspector or data recorder, the type of observation, 
the results, the acceptability and the action taken in connection with 
any deficiencies noted. The records must be retained for three years 
after the life of the packaging to which they apply.


Sec. 71.93  Inspection and tests.

    (a) The licensee or certificate holder shall permit the Commission, 
at all reasonable times, to inspect the licensed material, packaging, 
premises, and facilities in which the licensed material or packaging is 
used, provided, constructed, fabricated, tested, stored, or shipped.
    (b) The licensee shall perform, and permit the Commission to 
perform, any tests the Commission deems necessary or appropriate for 
the administration of the regulations in this chapter.
    (c) The licensee shall notify the Administrator of the appropriate 
NRC Regional Office listed in appendix A of part 73 of this chapter, at 
least 45 days before fabrication of a package to be used for the 
shipment of licensed material having a decay heat load in excess of 5 
kW or with a maximum normal operating pressure in excess of 103 kPa (15 
lbf/in2) gauge.


Sec. 71.95  Reports.

    The licensee shall report to the Director, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, within 30 days--
    (a) Any instance in which there is significant reduction in the 
effectiveness of any approved Type B, or fissile, packaging during use;
    (b) Details of any defects with safety significance in Type B, or 
fissile, packaging after first use, with the means employed to repair 
the defects and prevent their recurrence; or
    (c) Instances in which the conditions of approval in the 
certificate of compliance were not observed in making a shipment.


Sec. 71.97  Advance notification of shipment of irradiated reactor fuel 
and nuclear waste.

    (a) As specified in paragraphs (b), (c) and (d) of this section, 
each licensee shall provide advance notification to the governor of a 
State, or the governor's designee, of the shipment of licensed 
material, through, or across the boundary of the State, before the 
transport, or delivery to a carrier, for transport, of licensed 
material outside the confines of the licensee's plant or other place of 
use or storage.
    (b) Advance notification is required under this section for 
shipments of irradiated reactor fuel in quantities less than that 
subject to advance notification requirements of Sec. 73.37(f) of this 
chapter. Advance notification is also required under this section for 
shipment of licensed material, other than irradiated fuel, meeting the 
following three conditions:
    (1) The licensed material is required by this part to be in Type B 
packaging for transportation;
    (2) The licensed material is being transported to or across a State 
boundary en route to a disposal facility or to a collection point for 
transport to a disposal facility; and
    (3) The quantity of licensed material in a single package exceeds 
the least of the following:
    (i) 3000 times the A1 value of the radionuclides as specified 
in appendix A, Table A-1 for special form radioactive material;
    (ii) 3000 times the A2 value of the radionuclides as specified 
in appendix A, Table A-1 for normal form radioactive material; or
    (iii) 1000 TBq (27,000 Ci).
    (c) Procedures for submitting advance notification.
    (1) The notification must be made in writing to the office of each 
appropriate governor or governor's designee and to the Administrator of 
the appropriate 

[[Page 50279]]
NRC Regional Office listed in appendix A to part 73 of this chapter.
    (2) A notification delivered by mail must be postmarked at least 7 
days before the beginning of the 7-day period during which departure of 
the shipment is estimated to occur.
    (3) A notification delivered by messenger must reach the office of 
the governor or of the governor's designee at least 4 days before the 
beginning of the 7-day period during which departure of the shipment is 
estimated to occur.
    (i) A list of the names and mailing addresses of the governors' 
designees receiving advance notification of transportation of nuclear 
waste was published in the Federal Register on June 30, 1995 (60 FR 
34306).
    (ii) The list will be published annually in the Federal Register on 
or about June 30 to reflect any changes in information.
    (iii) A list of the names and mailing addresses of the governors' 
designees is available on request from the Director, Office of State 
Programs, U.S. Nuclear Regulatory Commission, Washington, DC 20555-
0001.
    (4) The licensee shall retain a copy of the notification as a 
record for 3 years.
    (d) Information to be furnished in advance notification of 
shipment. Each advance notification of shipment of irradiated reactor 
fuel or nuclear waste must contain the following information:
    (1) The name, address, and telephone number of the shipper, 
carrier, and receiver of the irradiated reactor fuel or nuclear waste 
shipment;
    (2) A description of the irradiated reactor fuel or nuclear waste 
contained in the shipment, as specified in the regulations of DOT in 49 
CFR 172.202 and 172.203(d);
    (3) The point of origin of the shipment and the 7-day period during 
which departure of the shipment is estimated to occur;
    (4) The 7-day period during which arrival of the shipment at State 
boundaries is estimated to occur;
    (5) The destination of the shipment, and the 7-day period during 
which arrival of the shipment is estimated to occur; and
    (6) A point of contact, with a telephone number, for current 
shipment information.
    (e) Revision notice. A licensee who finds that schedule information 
previously furnished to a governor or governor's designee, in 
accordance with this section, will not be met, shall telephone a 
responsible individual in the office of the governor of the State or of 
the governor's designee and inform that individual of the extent of the 
delay beyond the schedule originally reported. The licensee shall 
maintain a record of the name of the individual contacted for 3 years.
    (f) Cancellation notice.
    (1) Each licensee who cancels an irradiated reactor fuel or nuclear 
waste shipment for which advance notification has been sent shall send 
a cancellation notice to the governor of each State or to the 
governor's designee previously notified, and to the Administrator of 
the appropriate NRC Regional Office listed in appendix A of part 73 of 
this chapter.
    (2) The licensee shall state in the notice that it is a 
cancellation and identify the advance notification that is being 
canceled. The licensee shall retain a copy of the notice as a record 
for 3 years.


Sec. 71.99  Violations.

    (a) The Commission may obtain an injunction or other court order to 
prevent a violation of the provisions of--
    (1) The Atomic Energy Act of 1954, as amended;
    (2) Title II of the Energy Reorganization Act of 1974, as amended; 
or (3) A regulation or order issued pursuant to those Acts.
    (b) The Commission may obtain a court order for the payment of a 
civil penalty imposed under section 234 of the Atomic Energy Act:
    (1) For violations of--
    (i) Sections 53, 57, 62, 63, 81, 82, 101, 103, 104, 107, or 109 of 
the Atomic Energy Act of 1954, as amended;
    (ii) Section 206 of the Energy Reorganization Act;
    (iii) Any rule, regulation, or order issued pursuant to the 
sections specified in paragraph (b)(1)(i) of this section; or
    (iv) Any term , condition, or limitation of any license issued 
under the sections specified in paragraph (b)(1)(i) of this section.
    (2) For any violation for which a license may be revoked under 
section 186 of the Atomic Energy Act of 1954, as amended.


Sec. 71.100  Criminal penalties.

    (a) Section 223 of the Atomic Energy Act of 1954, as amended, 
provides for criminal sanctions for willful violation of, attempted 
violation of, or conspiracy to violate, any regulation issued under 
sections 161b, 161i, or 161o of the Act. For purposes of section 223, 
all the regulations in part 71 are issued under one or more of sections 
161b, 161i, or 161o, except for the sections listed in paragraph (b) of 
this section.
    (b) The regulations in part 71 that are not issued under sections 
161b, 161i, or 161o for the purposes of section 223 are as follows: 
Secs. 71.0, 71.2, 71.4, 71.6, 71.7, 71.9, 71.10, 71.31, 71.33, 71.35, 
71.37, 71.38, 71.39, 71.41, 71.43, 71.45, 71.47, 71.51, 71.52, 71.53, 
71.55, 71.59, 71.65, 71.71, 71.73, 71.74, 71.75, 71.77, 71.99, and 
71.100.

Subpart H--Quality Assurance


Sec. 71.101  Quality assurance requirements.

    (a) Purpose. This subpart describes quality assurance requirements 
applying to design, purchase, fabrication, handling, shipping, storing, 
cleaning, assembly, inspection, testing, operation, maintenance, 
repair, and modification of components of packaging that are important 
to safety. As used in this subpart, ``quality assurance'' comprises all 
those planned and systematic actions necessary to provide adequate 
confidence that a system or component will perform satisfactorily in 
service. Quality assurance includes quality control, which comprises 
those quality assurance actions related to control of the physical 
characteristics and quality of the material or component to 
predetermined requirements.
    (b) Establishment of program. Each licensee shall establish, 
maintain, and execute a quality assurance program satisfying each of 
the applicable criteria of Secs. 71.101 through 71.137 and satisfying 
any specific provisions that are applicable to the licensee's 
activities including procurement of packaging. The licensee shall apply 
each of the applicable criteria in a graded approach, i.e., to an 
extent that is consistent with its importance to safety.
    (c) Approval of program. Before the use of any package for the 
shipment of licensed material subject to this subpart, each licensee 
shall obtain Commission approval of its quality assurance program. Each 
licensee shall 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, with the Director, Office of 
Nuclear Material Safety and Safeguards, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001.
    (d) Existing package designs. The provisions of this paragraph deal 
with packages that have been approved for use in accordance with this 
part before January 1, 1979, and which have been designed in accordance 
with the provisions of this part in effect at the time of application 
for package approval. Those packages will be accepted as having been 
designed in accordance with a quality assurance program that satisfies 
the provisions of paragraph (b) of this section. 

[[Page 50280]]

    (e) Existing packages. The provisions of this paragraph deal with 
packages that have been approved for use in accordance with this part 
before January 1, 1979; have been at least partially fabricated prior 
to that date; and for which the fabrication is in accordance with the 
provisions of this part in effect at the time of application for 
approval of package design. These packages will be accepted as having 
been fabricated and assembled in accordance with a quality assurance 
program that satisfies the provisions of paragraph (b) of this section.
    (f) Previously approved programs. A Commission-approved quality 
assurance program that satisfies the applicable criteria of Appendix B 
of Part 50 of this chapter, and that is established, maintained, and 
executed with regard to transport packages, will be accepted as 
satisfying the requirements of paragraph (b) of this section. Before 
first use, the licensee shall notify the Director, Office of Nuclear 
Material Safety and Safeguards, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, of its intent to apply its previously 
approved Appendix B program to transportation activities. The licensee 
shall identify the program by date of submittal to the Commission, 
Docket Number, and date of Commission approval.


Sec. 71.103  Quality assurance organization.

    (a) The licensee 3 shall be responsible for the establishment 
and execution of the quality assurance program. The licensee may 
delegate to others, such as contractors, agents, or consultants, the 
work of establishing and executing the quality assurance program, or 
any part of the quality assurance program, but shall retain 
responsibility for the program. The licensee shall clearly establish 
and delineate, in writing, the authority and duties of persons and 
organizations performing activities affecting the safety-related 
functions of structures, systems, and components. These activities 
include performing the functions associated with attaining quality 
objectives and the quality assurance functions.

    \3\ While the term ``licensee'' is used in these criteria, the 
requirements are applicable to whatever design, fabrication, 
assembly, and testing of the package is accomplished with respect to 
a package prior to the time a package approval is issued.
---------------------------------------------------------------------------

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


Sec. 71.105   Quality assurance program.

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


Sec. 71.107   Package design control.

    (a) The licensee shall establish measures to assure that applicable 
regulatory requirements and the package design, as specified in the 
license for those materials and components to which this section 
applies, are correctly translated into specifications, drawings, 
procedures, and instructions. These measures must include provisions to 
assure that appropriate quality standards are specified and included in 
design documents and that deviations from standards are controlled. 
Measures must be established for the selection and review for 
suitability of application of materials, parts, equipment, and 
processes that are essential to the safety-related functions of the 
materials, parts, and components of the packaging.

[[Page 50281]]

    (b) The licensee shall establish measures for the identification 
and control of design interfaces and for coordination among 
participating design organizations. These measures must include the 
establishment of written procedures, among participating design 
organizations, for the review, approval, release, distribution, and 
revision of documents involving design interfaces. The design control 
measures must provide for verifying or checking the adequacy of design, 
by methods such as design reviews, alternate or simplified 
calculational methods, or by a suitable testing program. For the 
verifying or checking process, the licensee shall designate individuals 
or groups other than those who were responsible for the original 
design, but who may be from the same organization. Where a test program 
is used to verify the adequacy of a specific design feature in lieu of 
other verifying or checking processes, the licensee shall include 
suitable qualification testing of a prototype or sample unit under the 
most adverse design conditions. The licensee shall apply design control 
measures to items such as the following:
    (1) Criticality physics, radiation shielding, stress, thermal, 
hydraulic, and accident analyses;
    (2) Compatibility of materials;
    (3) Accessibility for inservice inspection, maintenance, and 
repair;
    (4) Features to facilitate decontamination; and
    (5) Delineation of acceptance criteria for inspections and tests.
    (c) The licensee shall subject design changes, including field 
changes, to design control measures commensurate with those applied to 
the original design. Changes in the conditions specified in the package 
approval require NRC approval.


Sec. 71.109   Procurement document control.

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


Sec. 71.111   Instructions, procedures, and drawings.

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


Sec. 71.113   Document control.

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


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

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


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

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


Sec. 71.119   Control of special processes.

    The licensee shall establish measures to assure that special 
processes, including welding, heat treating, and nondestructive 
testing, are controlled and accomplished by qualified personnel using 
qualified procedures in accordance with applicable codes, standards, 
specifications, criteria, and other special requirements.


Sec. 71.121   Internal inspection.

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


Sec. 71.123   Test control.

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

[[Page 50282]]



Sec. 71.125   Control of measuring and test equipment.

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


Sec. 71.127   Handling, storage, and shipping control.

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


Sec. 71.129   Inspection, test, and operating status.

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


Sec. 71.131   Nonconforming materials, parts, or components.

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


Sec. 71.133   Corrective action.

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


Sec. 71.135   Quality assurance records.

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


Sec. 71.137   Audits.

    The licensee shall carry out a comprehensive system of planned and 
periodic audits, to verify compliance with all aspects of the quality 
assurance program, and to determine the effectiveness of the program. 
The audits must be performed in accordance with written procedures or 
checklists by appropriately trained personnel not having direct 
responsibilities in the areas being audited. Audited results must be 
documented and reviewed by management having responsibility in the area 
audited. Follow-up action, including reaudit of deficient areas, must 
be taken where indicated.

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

    I. Values of A1 and A2 for individual radionuclides, 
which are the bases for many activity limits elsewhere in these 
regulations are given in Table A-1. The curie (Ci) values specified 
are obtained by converting from the Terabecquerel (TBq) figure. The 
curie values are expressed to three significant figures to assure 
that the difference in the TBq and Ci quantities is one tenth of one 
percent or less. Where values of A1 or A2 are unlimited, 
it is for radiation control purposes only. For nuclear criticality 
safety, some materials are subject to controls placed on fissile 
material.
    II. For individual radionuclides whose identities are known, but 
which are not listed in Table A-1, the determination of the values 
of A1 and A2 requires Commission approval, except that the 
values of A1 and A2 in Table A-2 may be used without 
obtaining Commission approval.
    III. In the calculations of A1 and A2 for a 
radionuclide not in Table A-1, a single radioactive decay chain, in 
which radionuclides are present in their naturally occurring 
proportions, and in which no daughter nuclide has a half-life either 
longer than 10 days, or longer than that of the parent nuclide, 
shall be considered as a single radionuclide, and the activity to be 
taken into account, and the A1 or A2 value to be applied 
shall be those corresponding to the parent nuclide of that chain. In 
the case of radioactive decay chains in which any daughter nuclide 
has a half-life either longer than 10 days, or greater than that of 
the parent nuclide, the parent and those daughter nuclides shall be 
considered as mixtures of different nuclides.
    IV. For mixtures of radionuclides whose identities and 
respective activities are known, the following conditions apply:
    (a) For special form radioactive material, the maximum quantity 
transported in a Type A package:
[GRAPHIC][TIFF OMITTED]TR28SE95.001

    (b) For normal form radioactive material, the maximum quantity 
transported in a Type A package:

[[Page 50283]]
[GRAPHIC][TIFF OMITTED]TR28SE95.002


Where B(i) is the activity of radionuclide I and A1(i) and 
A2(i) are the A1 and A2 values for radionuclide I, 
respectively.
    Alternatively, an A1 value for mixtures of special form 
material may be determined as follows:
[GRAPHIC][TIFF OMITTED]TR28SE95.003

Where f(i) is the fraction of activity of nuclide I in the mixture 
and A1(i) is the appropriate A1 value for nuclide I.
    An A2 value for mixtures of normal form material may be 
determined as follows:
[GRAPHIC][TIFF OMITTED]TR28SE95.004

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

                                                     Table A-1.--A1 and A2 Values for Radionuclides                                                     
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                                                              Specific activity         
    Symbol of         Element and          A1 (TBq)             A1 (Ci)            A2 (TBq)           A2 (Ci)      -------------------------------------
   radionuclide      atomic number                                                                                       (TBq/g)             (Ci/g)     
--------------------------------------------------------------------------------------------------------------------------------------------------------
Ac-225...........  Actinium(89).....  0.6                 16.2                1 x 10-2           0.270              2.1 x 103          5.8 x 104        
Ac-227...........                     40                  1080                2 x 10-5           5.41 x 10-4        2.7                7.2 x 101        
Ac-228...........                     0.6                 16.2                0.4                10.8               8.4 x 104          2.2 x 106        
Ag-105...........  Silver(47).......  2                   54.1                2                  54.1               1.1 x 103          3.0 x 104        
Ag-108m..........                     0.6                 16.2                0.6                16.2               9.7 x 10-1         2.6 x 101        
Ag-110m..........                     0.4                 10.8                0.4                10.8               1.8 x 103          4.7 x 103        
Ag-111...........                     0.6                 16.2                0.5                13.5               5.8 x 103          1.6 x 105        
Al-26............  Aluminum(13).....  0.4                 10.8                0.4                10.8               7.0 x 10-4         1.9 x 10-2       
Am-241...........  Americium(95)....  2                   54.1                2 x 10-4           5.41 x 10-3        1.3 x 10-1         3.4              
Am-242m..........                     2                   54.1                2 x 10-4           5.41 x 10-3        3.6 x 10-1         9.7 x 105        
Am-243...........                     2                   54.1                2 x 10-4           5.41 x 10-3        7.4 x 10-3         2.0 x 10-1       
Ar-37............  Argon(18)........  40                  1080                40                 1080               3.7 x 103          9.9 x 104        
Ar-39............                     20                  541                 20                 541                1.3 x 103          3.4 x 101        
Ar-41............                     0.6                 16.2                0.6                16.2               1.5 x 106          4.2 x 107        
Ar-42............                     0.2                 5.41                0.2                5.41               9.6                2.6 x 102        
As-72............  Arsenic(33)......  0.2                 5.41                0.2                5.41               6.2 x 104          1.7 x 106        
As-73............                     40                  1080                40                 1080               8.2 x 102          2.2 x 104        
As-74............                     1                   27.0                0.5                13.5               3.7 x 103          9.9 x 104        
As-76............                     0.2                 5.41                0.2                5.41               5.8 x 104          1.6 x 106        
As-77............                     20                  541                 0.5                13.5               3.9 x 104          1.0 x 106        
At-211...........  Astatine(85).....  30                  811                 2                  54.1               7.6 x 104          2.1 x 106        
Au-193...........  Gold(79).........  6                   162                 6                  162                3.4 x 104          9.2 x 105        
Au-194...........                     1                   27.0                1                  27.0               1.5 x 104          4.1 x 105        
Au-195...........                     10                  270                 10                 270                1.4 x 102          3.7 x 103        
Au-196...........                     2                   54.1                2                  54.1               4.0 x 103          1.1 x 105        
Au-198...........                     3                   81.1                0.5                13.5               9.0 x 103          2.4 x 105        
Au-199...........                     10                  270                 0.9                24.3               7.7 x 103          2.1 x 105        
Ba-131...........  Barium(56).......  2                   54.1                2                  54.1               3.1 x 103          8.4 x 104        
Ba-133m..........                     10                  270                 0.9                24.3               2.2 x 104          6.1 x 105        
Ba-133...........                     3                   81.1                3                  81.1               9.4                2.6 x 102        
Ba-140...........                     0.4                 10.8                0.4                10.8               2.7 x 103          7.3 x 104        
Be-7.............  Beryllium(4).....  20                  541                 20                 541                1.3 x 104          3.5 x 105        
Be-10............                     20                  541                 0.5                13.5               8.3 x 10-4          2.2 x 10-2      
Bi-205...........  Bismuth(83)......  0.6                 16.2                0.6                16.2               1.5 x 10-3         4.2 x 104        
Bi-206...........                     0.3                 8.11                0.3                8.11               3.8 x 103          1.0 x 105        

[[Page 50284]]
                                                                                                                                                        
Bi-207...........                     0.7                 18.9                0.7                18.9               1.9                5.2 x 101        
Bi-210m..........                     0.3                 8.11                3 x 10-2           0.811              2.1 x 10-5         5.7 x 10-4       
Bi-210...........                     0.6                 16.2                0.5                13.5               4.6 x 103          1.2 x 105        
Bi-212...........                     0.3                 8.11                0.3                8.11               5.4 x 105          1.5 x 107        
Bk-247...........  Berkelium(97)....  2                   54.1                2 x 10-4           5.41 x 10-3        3.8 x 10-2         1.0              
Bk-249...........                     40                  1080                8 x 10-2           2.16               6.1 x 101          1.6 x 103        
Br-76............  Bromine(35)......  0.3                 8.11                0.3                8.11               9.4 x 104          2.5 x 106        
Br-77............                     3                   81.1                3                  81.1               2.6 x 104          7.1 x 105        
Br-82............                     0.4                 108                 0.4                10.8               4.0 x 104          1.1 x 106        
C-11.............  Carbon(6)........  1                   270                 0.5                13.5               3.1 x 107          8.4 x 108        
C-14.............                     40                  1080                2                  54.1               1.6 x 10-1         4.5              
Ca-41............  Calcium(20)......  40                  1080                40                 1080               3.1 x 10-3         8.5 x 10-2       
Ca-45............                     40                  1080                0.9                24.3               6.6 x 102          1.8 x 104        
Ca-47............                     0.9                 24.3                0.5                13.5               2.3 x 104          6.1 x 105        
Cd-109...........  Cadmium(48)......  40                  1080                1                  27.0               9.6 x 101          2.6 x 103        
Cd-113m..........                     20                  541                 9 x 10-2           2.43               8.3 x 104          2.2 x 102        
Cd-115m..........                     0.3                 8.11                0.3                8.11               9.4 x 102          2.5 x 104        
Cd-115...........                     4                   108                 0.5                13.5               1.9 x 104          5.1 x 105        
Ce-139...........  Cerium(58).......  6                   162                 6                  162                2.5 x 102          6.8 x 103        
Ce-141...........                     10                  270                 0.5                13.5               1.1 x 103          2.8 x 104        
Ce-143...........                     0.6                 16.2                0.5                13.5               2.5 x 104          6.6 x 105        
Ce-144...........                     0.2                 5.41                0.2                5.41               1.2 x 102          3.2 x 103        
Cf-248...........  Californium(98)..  30                  811                 3 x 10-3           8.11 x 10-2        5.8 x 101          1.6 x 103        
Cf-249...........                     2                   54.1                2 x 10-4           5.41 x 10-3        1.5 x 10-1         4.1              
Cf-250...........                     5                   135                 5 x 10-4           1.35 x 10-2        4.0                1.1 x 102        
Cf-251...........                     2                   54.1                2 x 10-4           5.41 x 10-3        5.9 x 10-2         1.6              
Cf-252...........                     0.1                 2.70                1 x 10-3           2.70 x 10-2        2.0 x 101          5.4 x 102        
Cf-253...........                     40                  1080                6 x 10-2           1.62               1.1 x 103          2.9 x 104        
Cf-254...........                     3 x 10-3            8.11 x 10-2         6 x 10-4           1.62 x 10-2        3.1 x 102          8.5 x 103        
Cl-36............  Chlorine(17).....  20                  541                 0.5                13.5               1.2 x 10-3         3.3 x 10-2       
Cl-38............                     0.2                 5.41                0.2                5.41               4.9 x 106          1.3 x 108        
Cm-240...........  Curium(96).......  40                  1080                2 x 10-2           0.541              7.5 x 102          2.0 x 104        
Cm-241...........                     2                   54.1                0.9                24.3               6.1 x 102          1.7 x 104        
Cm-242...........                     40                  1080                1 x 10-2           0.270              1.2 x 102          3.3 x 103        
Cm-243...........                     3                   81.1                3 x 10-4           8.11 x 10-3        1.9                5.2 x 101        
Cm-244...........                     4                   1080                4 x 10-4           1.08 x 10-2        3.0                8.1 x 105        
Cm-245...........                     2                   54.1                2 x 10-4           5.41 x 10-3        6.4 x 10-3         1.7 x 10-1       
Cm-246...........                     2                   54.1                2 x 10-4           5.41 x 10-3        1.1 x 10-2         3.1 x 10-1       
Cm-247...........                     2                   54.1                2 x 10-4           5.41 x 10-3        3.4 x 10-6         9.3 x 10-5       
Cm-248...........                     4 x 10-2            1.08                5 x 10-5           1.35 x 10-3        1.6 x 10-4         4.2 x 10-3       
Co-55............  Cobalt(27).......  0.5                 13.5                0.5                13.5               1.1 x 105          3.1 x 106        
Co-56............                     0.3                 8.11                0.3                8.11               1.1 x 103          3.0 x 104        
Co-57............                     8                   216                 8                  216                3.1 x 102          8.4 x 103        
Co-58m...........                     40                  1080                40                 1080               2.2 x 105          5.9 x 106        
Co-58............                     1                   27.0                1                  27.0               1.2 x 103          3.2 x 104        
Co-60............                     0.4                 10.8                0.4                10.8               4.2 x 101          1.1 x 103        
Cr-51............  Chromium(24).....  30                  811                 30                 811                3.4 x 103          9.2 x 104        
Cs-129...........  Cesium(55).......  4                   108                 4                  108                2.8 x 104          7.6 x 105        
Cs-131...........                     40                  1080                40                 1080               3.8 x 103          1.0 x 105        
Cs-132...........                     1                   27.0                1                  27.0               5.7 x 103          1.5 x 105        
Cs-134m..........                     40                  1080                9                  243                3.0 x 105          8.0 x 106        
Cs-134...........                     0.6                 16.2                0.5                13.5               4.8 x 101          1.3 x 103        
Cs-135...........                     40                  1080                0.9                24.3               4.3 x 10-5         1.2 x 10-3       
Cs-136...........                     0.5                 13.5                0.5                13.5               2.7 x 103          7.3 x 104        
Cs-137...........                     2                   54.1                0.5                13.5               3.2                8.7 x 101        
Cu-64............  Copper(29).......  5                   135                 0.9                24.3               1.4 x 105          3.9 x 106        
Cu-67............                     9                   243                 0.9                24.3               2.8 x 104          7.6 x 105        
Dy-159...........  Dysprosium(66)...  20                  541                 20                 541                2.1 x 102          5.7 x 103        
Dy-165...........                     0.6                 16.2                0.5                13.5               3.0 x 105          8.2 x 106        
Dy-166...........                     0.3                 8.11                0.3                8.11               8.6 x 103          2.3 x 105        
Er-169...........  Erbium(68).......  40                  1080                0.9                24.3               3.1 x 103          8.3 x 104        
Er-171...........                     0.6                 16.2                0.5                13.5               9.0 x 104          2.4 x 106        
Es-253...........  Einsteinium(99) a  40                  1080                5 x 10-1           1.35               .................  .................
Es-254...........                     30                  811                 3 x 10-3           8.11 x 10-2        .................  .................
Es-254m..........                     0.6                 16.2                0.4                10.8               .................  .................
Es-255...........                                                                                                                                       
Eu-147...........  Europium(63).....  2                   54.1                2                  54.1               1.4 x 103          3.7 x 104        
Eu-148...........                     0.5                 13.5                0.5                13.5               6.0 x 102          1.6 x 104        
Eu-149...........                     20                  541                 20                 541                3.5 x 102          9.4 x 103        
Eu-150...........                     0.7                 18.9                0.7                18.9               6.1 x 104          6.7 x 106        

[[Page 50285]]
                                                                                                                                                        
Eu-152m..........                     0.6                 16.2                0.5                13.5               8.2 x 104          2.2 x 106        
Eu-152...........                     0.9                 24.3                0.9                24.3               6.5                1.8 x 102        
Eu-154...........                     0.8                 21.6                0.5                13.5               9.8                2.6 x 102        
Eu-155...........                     20                  541                 2                  54.1               1.8 x 101          4.9 x 103        
Eu-156...........                     0.6                 16.2                0.5                13.5               2.0 x 103          5.5 x 104        
F-18.............  Fluorine(9)......  1                   27.0                0.5                13.5               3.5 x 105          9.5 x 107        
Fe-52............  Iron(26).........  0.2                 5.41                0.2                5.41               2.7 x 105          7.3 x 106        
Fe-55............                     40                  1080                40                 1080               8.8 x 101          2.4 x 103        
Fe-59............                     0.8                 21.6                0.8                21.6               1.8 x 103          3.0 x 10-4       
Fe-60............                     40                  1080                0.2                5.41               7.4 x 10-4         2.0 x 10-2       
Fm-255...........  Fermium(100) b...  40                  1080                0.8                21.6                                                   
Fm-257...........                     40                  1080                7 x 10-3           1.89 x 10-3                                            
Ga-67............  Gallium(31)......  6                   162                 6                  162                2.2 x 104          6.0 x 105        
Ga-68............                     0.3                 8.11                0.3                8.11               1.5 x 106          4.1 x 107        
Ga-72............                     0.4                 10.8                0.4                10.8               1.1 x 105          3.1 x 106        
Gd-146...........  Gadolinium(64)...  0.4                 10.8                0.4                10.8               6.9 x 102          1.9 x 104        
Gd-148...........                     3                   81.1                3 x 10-4           8.11 x 10-3        6.7                2.9 x 101        
Gd-153...........                     10                  270                 5                  135                1.3 x 102          3.5 x 103        
Gd-159...........                     4                   108                 0.5                13.5               3.9 x 104          1.1 x 106        
Ge-68............  Germanium(32)....  0.3                 8.11                0.3                8.11               2.6 x 102          7.1 x 103        
Ge-71............                     40                  1080                40                 1080               5.8 x 103          1.6 x 105        
Ge-77............                     0.3                 8.11                0.3                8.11               1.3 x 105          3.6 x 106        
H-3..............  Hydrogen(1)......  See T-Tritium                                                                                                     
Hf-172...........  Hafnium(72)......  0.5                 13.5                0.3                8.11               4.1 x 101          1.1 x 103        
Hf-175...........                     3                   81.1                3                  81.1               3.9 x 102          1.1 x 104        
Hf-181...........                     2                   54.1                0.9                24.3               6.3 x 102          1.7 x 104        
Hf-182...........                     4                   108                 3 x 10-2           0.811              8.1 x 10-6         2.2 x 10-4       
Hg-194...........  Mercury(80)......  1                   27.0                1                  27.0               1.3 x 10-1         3.5              
Hg-195m..........                     5                   135                 5                  135                1.5 x 104          4.0 x 105        
Hg-197m..........                     10                  270                 0.9                24.3               2.5 x 104          6.7 x 105        
Hg-197...........                     10                  270                 10                 270                9.2 x 103          2.5 x 105        
Hg-203...........                     4                   108                 0.9                24.3               5.1 x 102          1.4 x 104        
Ho-163...........  Holmium(67)......  40                  1080                40                 1080               2.7                7.6 x 101        
Ho-166m..........                     0.6                 16.2                0.3                8.11               6.6 x 10-2         1.8              
Ho-166...........                     0.3                 8.11                0.3                8.11               2.6 x 104          7.0 x 105        
I-123............  Iodine(53).......  6                   162                 6                  162                7.1 x 104          1.9 x 106        
I-124............                     0.9                 24.3                0.9                24.3               9.3 x 103          2.5 x 105        
I-125............                     20                  541                 2                  54.1               6.4 x 102          1.7 x 104        
I-126............                     2                   54.1                0.9                24.3               2.9 x 103          8.0 x 104        
I-129............                     Unlimited           Unlimited           Unlimited          Unlimited          6.5 x 10-6         1.8 x 10-4       
I-131............                     3                   81.1                0.5                13.5               4.6 x 103          1.2 x 105        
I-132............                     0.4                 10.8                0.4                10.8               3.8 x 105          1.0 x 107        
I-133............                     0.6                 16.2                0.5                13.5               4.2 x 104          1.1 x 106        
I-134............                     0.3                 8.11                0.3                8.11               9.9 x 105          2.7 x 107        
I-135............                     0.6                 16.2                0.5                13.5               1.3 x 105          3.5 x 106        
In-111...........  Indium(49).......  2                   54.1                2                  54.1               1.5 x 104          4.2 x 105        
In-113m..........                     4                   108                 4                  108                6.2 x 105          1.7 x 107        
In-114m..........                     0.3                 8.11                0.3                8.11               8.6 x 102          2.3 x 104        
In-115m..........                     6                   162                 0.9                24.3               2.2 x 105          6.1 x 106        
Ir-189...........  Iridium(77)......  10                  270                 10                 270                1.9 x 103          5.2 x 104        
Ir-190...........                     0.7                 18.9                0.7                18.9               2.3 x 103          6.2 x 104        
Ir-192...........                     1                   27.0                0.5                13.5               3.4 x 102          9.2 x 103        
Ir-193m..........                     10                  270                 10                 270                2.4 x 103          6.4 x 104        
Ir-194...........                     0.2                 5.41                0.2                5.41               3.1 x 104          8.4 x 105        
K-40.............  Potassium(19)....  0.6                 16.2                0.6                16.2               2.4 x 10-7         6.4 x 10-6       
K-42.............                     0.2                 5.41                0.2                5.41               2.2 x 105          6.0 x 106        
K-43.............                     1.0                 27.0                0.5                13.5               1.2 x 105          3.3 x 106        
Kr-81............  Krypton(36)......  40                  1080                40                 1080               7.8 x 10-4         2.1 x 10-2       
Kr-85m...........                     6                   162                 6                  162                3.0 x 105          8.2 x 106        
Kr-85............                     20                  541                 10                 270                1.5 x 101          3.9 x 102        
Kr-87............                     0.2                 5.41                0.2                5.41               1.0 x 106          2.8 x 107        
La-137...........  Lanthanum(57)....  40                  1080                2                  54.1               1.6 x 10-3         4.4 x 10-2       
La-140...........                     0.4                 10.8                0.4                10.8               2.1 x 104          5.6 x 105        
Lu-172...........  Lutetium(71).....  0.5                 13.5                0.5                13.5               4.2 x 103          1.1 x 105        
Lu-173...........                     8                   216                 8                  216                5.6 x 101          1.5 x 103        
Lu-174m..........                     20                  541                 8                  216                2.0 x 102          5.3 x 103        
Lu-174...........                     8                   216                 4                  108                2.3 x 101          6.2 x 102        
Lu-177...........                     30                  811                 0.9                24.3               4.1 x 103          1.1 x 105        
MFP..............                                                                                                                                       
(6)For mied                                                                                                                                             
 fission                                                                                                                                                
 products, use                                                                                                                                          
 formula for                                                                                                                                            
 mitures or Table                                                                                                                                       
 A-2                                                                                                                                                    
Mg-28............  Magnesium(12)....  0.2                 5.41                0.2                5.41               2.0 x 105          5.4 x 106        

[[Page 50286]]
                                                                                                                                                        
Mn-52............  Manganese(25)....  0.3                 8.11                0.3                8.11               1.6 x 104          4.4 x 105        
Mn-53............                     Unlimited           Unlimited           Unlimited          Unlimited          6.8 x 10-5         1.8 x 10-3       
Mn-54............                     1                   27.0                1                  27.0               2.9 x 102          7.7 x 103        
Mn-56............                     0.2                 5.41                0.2                5.41               8.0 x 105          2.2 x 107        
Mo-93............  Molybdenum(42)...  40                  1080                7                  189                4.1 x 10-2         1.1              
Mo-99............                     0.6                 16.2                0.5                13.5c              1.8 x 104          4.8 x 105        
N-13.............  Nitrogen(7)......  0.6                 16.2                0.5                13.5               5.4 x 107          1.5 x 109        
Na-22............  Sodium(11).......  0.5                 13.5                0.5                13.5               2.3 x 102          6.3 x 103        
Na-24............                     0.2                 5.41                0.2                5.41               3.2 x 105          8.7 x 106        
Nb-92m...........  Niobium(41)......  0.7                 18.9                0.7                18.9               5.2 x 103          1.4 x 105        
Nb-93m...........                     40                  1080                6                  162                8.8                2.4 x 102        
Nb-94............                     0.6                 16.2                0.6                16.2               6.9 x 10-3         1.9 x 10-1       
Nb-95............                     1                   27.0                1                  27.0               1.5 x 103          3.9 x 104        
Nb-97............                     0.6                 16.2                0.5                13.5               9.9 x 105          2.7 x 107        
Nd-147...........  Neodymium(60)....  4                   108                 0.5                13.5               3.0 x 103          8.1 x 104        
Nd-149...........                     0.6                 16.2                0.5                13.5               4.5 x 105          1.2 x 107        
Ni-59............  Nickel(28).......  40                  1080                40                 1080               3.0 x 10-3         8.0 x 10-2       
Ni-63............                     40                  1080                30                 811                2.1                5.7 x 101        
Ni-65............                     0.3                 8.11                0.3                8.11               7.1 x 105          1.9 x 107        
Np-235...........  Neptunium(93)....  40                  1080                40                 1080               5.2 x 101          1.4 x 103        
Np-236...........                     7                   189                 1 x 10-3           2.70 x 10-2        4.710-4-4          1.3 x 10-2       
Np-237...........                     2                   54.1                2 x 10-4           5.41 x 10-3        2.6 x 10-5         7.1 x 10-4       
Np-239...........                     6                   162                 0.5                13.5               8.6 x 103          2.3 x 105        
Os-185...........  Osmium(76).......  1                   27.0                1                  27.0               2.8 x 102          7.5 x 103        
Os-191m..........                     40                  1080                40                 1080               4.6 x 104          1.3 x 106        
Os-191...........                     10                  270                 0.9                24.3               1.6 x 103          4.4 x 104        
Os-193...........                     0.6                 16.2                0.5                13.5               2.0 x 104          5.3 x 105        
Os-194...........                     0.2                 5.41                0.2                5.41               1.1 x 101          3.1 x 102        
P-32.............  Phosphorus(15)...  0.3                 8.11                0.3                8.11               1.1 x 104          2.9 x 105        
P-33.............                     40                  1080                0.9                24.3               5.8 x 103          1.6 x 105        
Pa-230...........  Protactinium(91).  2                   54.1                0.1                2.70               1.2 x 103          3.3 x 104        
Pa-231...........                     0.6                 16.2                6 x 10-5           1.62 x 10-3        1.7 x 10-3         4.7 x 10-2       
Pa-233...........                     5                   135                 0.9                24.3               7.7 x 102          2.1 x 104        
Pb-201...........  Lead(82).........  1                   27.0                1                  27.0               6.2 x 104          1.7 x 106        
Pb-202...........                     40                  1080                2                  54.1               1.2 x 10-4         3.4 x 10-3       
Pb-203...........                     3                   81.1                3                  81.1               1.1 x 104          3.0 x 105        
Pb-205...........                     Unlimited           Unlimited           Unlimited          Unlimited          4.5 x 10-6         1.2 x 10-4       
Pb-210...........                     0.6                 16.2                9 x 10-3           0.243              2.8                7.6 x 101        
Pb-212...........                     0.3                 8.11                0.3                8.11               5.1 x 104          1.4 x 106        
Pd-103...........  Palladium(46)....  40                  1080                40                 1080               2.8 x 103          7.5 x 104        
Pd-107...........                     Unlimited           Unlimited           Unlimited          Unlimited          1.9 x 10-5         5.1 x 10-4       
Pd-109...........                     0.6                 16.2                0.5                13.5               7.9 x 104          2.1 x 106        
Pm-143...........  Promethium(61)...  3                   81.1                3                  81.1               1.3 x 102          3.4 x 103        
Pm-144...........                     0.6                 16.2                0.6                16.2               9.2 x 101          2.5 x 103        
Pm-145...........                     30                  811                 7                  189                5.2                1.4 x 102        
Pm-147...........                     40                  1080                0.9                24.3               3.4 x 101          9.3 x 102        
Pm-148m..........                     0.5                 13.5                0.5                13.5               7.9 x 102          2.1 x 104        
Pm-149...........                     0.6                 16.2                0.5                13.5               1.5 x 104          4.0 x 105        
Pm-151...........                     3                   81.1                0.5                13.5               2.7 x 104          7.3 x 105        
Po-208...........  Polonium(84).....  40                  1080                2 x 10-2           0.541              2.2 x 101          5.9 x 102        
Po-209...........                     40                  1080                2 x 10-2           0.541              6.2 x 10-1         1.7 x 101        
Po-210...........                     40                  1080                2 x 10-2           0.541              1.7 x 102          4.5 x 103        
Pr-142...........  Praseodymium(59).  0.2                 5.41                0.2                5.41               4.3 x 104          1.2 x 106        
Pr-143...........                     4                   108                 0.5                13.5               2.5 x 103          6.7 x 104        
Pt-188...........  Platinum(78).....  0.6                 16.2                0.6                16.2               2.5 x 103          6.8 x 104        
Pt-191...........                     3                   81.1                3                  81.1               8.7 x 103          2.4 x 105        
Pt-193m..........                     40                  1080                9                  243                5.8 x 103          1.6 x 105        
Pt-193...........                     40                  1080                40                 1080               1.4                3.7 x 101        
Pt-195m..........                     10                  270                 2                  54.1               6.2 x 103          1.7 x 105        
Pt-197m..........                     10                  270                 0.9                24.3               3.4 x 105          1.0 x 107        
Pt-197...........                     20                  541                 0.5                13.5               3.2 x 104          8.7 x 105        
Pu-236...........  Plutonium(94)....  7                   189                 7 x 10-4           1.89 x 10-2        2.0 x 101          5.3 x 102        
Pu-237...........                     20                  541                 20                 541                4.5 x 102          1.2 x 104        
Pu-238...........                     2                   54.1                2 x 10-4           5.41 x 10-3        6.3 x 10-1         1.7 x 101        
Pu-239...........                     2                   54.1                2 x 10-4           5.41 x 10-3        2.3 x 10-3         6.2 x 10-2       
Pu-240...........                     2                   54.1                2 x 10-4           5.41 x 10-3        8.4 x 10-3         2.3 x 10-1       
Pu-241...........                     40                  1080                1 x 10-2           0.270              3.8                1.0 x 102        
Pu-242...........                     2                   54.1                2 x 10-4           5.41 x 10-3        1.5 x 10-4         3.9 x 10-3       
Pu-244...........                     0.3                 8.11                2 x 10-4           5.41 x 10-3        6.7 x 10-7         1.8 x 10-5       
Ra-223...........  Radium(88).......  0.6                 16.2                3 x 10-2           0.811              1.9 x 103          5.1 x 104        

[[Page 50287]]
                                                                                                                                                        
Ra-224...........                     0.3                 8.11                6 x 10-2           1.62               5.9 x 103          1.6 x 105        
Ra-225...........                     0.6                 16.2                2 x 10-2           0.541              1.5 x 103          3.9 x 104        
Ra-226...........                     0.3                 8.11                2 x 10-2           0.541              3.7 x 10-2         1.0              
Ra-228...........                     0.6                 16.2                4 x 10-2           1.08               1.0 x 101          2.7 x 102        
Rb-81............  Rubidium(37).....  2                   54.1                0.9                24.3               3.1 x 105          8.4 x 106        
Rb-83............                     2                   54.1                2                  54.1               6.8 x 102          1.8 x 104        
Rb-84............                     1                   27.0                0.9                24.3               1.8 x 103          4.7 x 104        
Rb-86............                     0.3                 8.11                0.3                8.11               3.0 x 103          8.1 x 104        
Rb-87............                     Unlimited           Unlimited           Unlimited          Unlimited          3.2 x 10-9         8.6 x 10-8       
Rb (natural).....                     Unlimited           Unlimited           Unlimited          Unlimited          6.7 x 106          1.8 x 108        
Re-183...........  Rhenium(75)......  5                   135                 5                  135                3.8 x 102          1.0 x 104        
Re-184m..........                     3                   81.1                3                  81.1               1.6 x 102          4.3 x 103        
Re-184...........                     1                   27.0                1                  27.0               6.9 x 102          1.9 x 104        
Re-186...........                     4                   108                 0.5                13.5               6.9 x 103          1.9 x 105        
Re-187...........                     Unlimited           Unlimited           Unlimited          Unlimited          1.4 x 10-9         3.8 x 10-8       
Re-188...........                     0.2                 5.41                0.2                5.41               3.6 x 104          9.8 x 105        
Re-189...........                     4                   108                 0.5                13.5               2.5 x 104          6.8 x 105        
Re (natural).....                     Unlimited           Unlimited           Unlimited          Unlimited                             2.4 x 10-8       
Rh-99............  Rhodium(45)......  2                   54.1                2                  54.1               3.0 x 103          8.2 x 104        
Rh-101...........                     4                   108                 4                  108                4.1 x 101          1.1 x 103        
Rh-102m..........                     2                   54.1                0.9                24.3               2.3 x 102          6.2 x 103        
Rh-102...........                     0.5                 13.5                0.5                13.5               4.5 x 101          1.2 x 103        
Rh-103m..........                     40                  1080                40                 1080               1.2 x 106          3.3 x 107        
Rh-105...........                     10                  270                 0.9                24.3               3.1 x 104          8.4 x 105        
Rn-222...........  Radon(86)........  0.2                 5.41                4 x 10-3           0.108              5.7 x 103          1.5 x 105        
Ru-97............  Ruthenium(44)....  4                   108                 4                  108                1.7 x 104          4.6 x 105        
Ru-103...........                     2                   54.1                0.9                24.3               1.2 x 103          3.2 x 104        
Ru-105...........                     0.6                 16.2                0.5                13.5               2.5 x 105          6.7 x 106        
Ru-106...........                     0.2                 5.41                0.2                5.41               1.2 x 102          3.3 x 103        
S-35.............  Sulfur(16).......  40                  1080                2                  54.1               1.6 x 103          4.3 x 104        
Sb-122...........  Antimony(51).....  0.3                 8.11                0.3                8.11               1.5 x 104          4.0 x 105        
Sb-124...........                     0.6                 16.2                0.5                13.5               6.5 x 102          1.7 x 104        
Sb-125...........                     2                   54.1                0.9                24.3               3.9 x 101          1.0 x 103        
Sb-126...........                     0.4                 10.8                0.4                10.8               3.1 x 103          8.4 x 104        
Sc-44............  Scandium(21).....  0.5                 13.5                0.5                13.5               6.7 x 105          1.8 x 107        
Sc-46............                     0.5                 13.5                0.5                13.5               1.3 x 103          3.4 x 104        
Sc-47............                     9                   243                 0.9                24.3               3.1 x 104          8.3 x 105        
Sc-48............                     0.3                 8.11                0.3                8.11               5.5 x 104          1.5 x 106        
Se-75............  Selenium(34).....  3                   81.1                3                  81.1               5.4 x 102          1.5 x 104        
Se-79............                     40                  1080                2                  54.1               2.6 x 10-3         7.0 x 10-2       
Si-31............  Silicon(14)......  0.6                 16.2                0.5                13.5               1.4 x 106          3.9 x 107        
Si-32............                     40                  1080                0.2                5.41               3.9                1.1 x 102        
Sm-145...........  Samarium(62).....  20                  541                 20                 541                9.8 x 101          2.6 x 103        
Sm-147...........                     Unlimited           Unlimited           Unlimited          Unlimited          8.5 x 10-1         2.3 x 10-8       
Sm-151...........                     40                  1080                4                  108                9.7 x 10-1         2.6 x 101        
Sm-153...........                     4                   108                 0.5                13.5               1.6 x 104          4.4 x 105        
Sn-113...........  Tin(50)..........  4                   108                 4                  108                3.7 x 102          1.0 x 104        
Sn-117m..........                     6                   162                 2                  54.1               3.0 x 103          8.2 x 104        
Sn-119m..........                     40                  1080                40                 1080               1.4 x 102          3.7 x 103        
Sn-121m..........                     40                  1080                0.9                24.3               2.0                5.4 x 101        
Sn-123...........                     0.6                 16.2                0.5                13.5               3.0 x 102          8.2 x 103        
Sn-125...........                     0.2                 5.41                0.2                5.41               4.0 x 103          1.1 x 105        
Sn-126...........                     0.3                 8.11                0.3                8.11               1.0 x 10-3         2.8 x 10-2       
Sr-82............  Strontium(38)....  0.2                 5.41                0.2                5.41               2.3 x 103          6.2 x 104        
Sr-85m...........                     5                   135                 5                  135                1.2 x 106          3.3 x 107        
Sr-85............                     2                   54.1                2                  54.1               8.8 x 102          2.4 x 104        
Sr-87m...........                     3                   81.1                3                  81.1               4.8 x 105          1.3 x 107        
Sr-89............                     0.6                 16.2                0.5                13.5               1.1 x 103          2.9 x 104        
Sr-90............                     0.2                 5.41                0.1                2.70               5.1                1.4 x 102        
Sr-91............                     0.3                 8.11                0.3                8.11               1.3 x 105          3.6 x 106        
Sr-92............                     0.8                 21.6                0.5                13.5               4.7 x 105          1.3 x 107        
T................  Tritium(1).......  40                  1080                40                 1080               3.6 x 102          9.7 x 103        
Ta-178...........  Tantalum(73).....  1                   27.0                1                  27.0               4.2 x 106          1.1 x 108        
Ta-179...........                     30                  811                 30                 811                4.1 x 101          1.1 x 103        
Ta-182...........                     0.8                 21.6                0.5                13.5               2.3 x 102          6.2 x 103        
Tb-157...........  Terbium(65)......  40                  1080                10                 270                5.6 x 10-1         1.5 x 101        
Tb-158...........                     1                   27.0                0.7                18.9               5.6 x 10-1         1.5 x 101        
Tb-160...........                     0.9                 24.3                0.5                13.5               4.2 x 102          1.1 x 104        
Tc-95m...........  Technetium(43)...  2                   54.1                2                  54.1               8.3 x 102          2.2 x 104        
Tc-96m...........                     0.4                 10.8                0.4                10.8               1.4 x 106          3.8 x 107        

[[Page 50288]]
                                                                                                                                                        
Tc-96............                     0.4                 10.8                0.4                10.8               1.2 x 104          3.2 x 105        
Tc-97m...........                     40                  1080                40                 1080               5.6 x 102          1.5 x 104        
Tc-97............                     Unlimited           Unlimited           Unlimited          Unlimited          5.2 x 10-5         1.4 x 10-3       
Tc-98............                     0.7                 18.9                0.7                18.9               3.2 x 10-5         8.7 x 10-4       
Tc-99m...........                     8                   216                 8                  216                1.9 x 105          5.3 x 106        
Tc-99............                     40                  1080                0.9                24.3               6.3 x 10-4         1.7 x 10-2       
Te-118...........  Tellurium(52)....  0.2                 5.41                0.2                5.41               6.8 x 103          1.8 x 105        
Te-121m..........                     5                   135                 5                  135                2.6 x 102          7.0 x 103        
Te-121...........                     2                   54.1                2                  54.1               2.4 x 103          6.4 x 104        
Te-123m..........                     7                   189                 7                  189                3.3 x 102          8.9 x 103        
Te-125m..........                     30                  811                 9                  243                6.7 x 102          1.8 x 104        
Te-127m..........                     20                  541                 0.5                13.5               3.5 x 102          9.4 x 103        
Te-127...........                     20                  541                 0.5                13.5               9.8 x 104          2.6 x 106        
Te-129m..........                     0.6                 16.2                0.5                13.5               1.1 x 103          3.0 x 104        
Te-129...........                     0.6                 16.2                0.5                13.5               7.7 x 105          2.1 x 107        
Te-131m..........                     0.7                 18.9                0.5                13.5               3.0 x 104          8.0 x 105        
Te-132...........                     0.4                 10.8                0.4                10.8               1.1 x 104          3.0 x 105        
Th-227...........  Thorium(90)......  9                   243                 1 x 10-2           0.270              1.1 x 103          3.1 x 104        
Th-228...........                     0.3                 8.11                4 x 10-4           1.08 x 10-2        3.0 x 101          8.2 x 102        
Th-229...........                     0.3                 8.11                3 x 10-5           8.11 x 10-4        7.9 x 10-3         2.1 x 10-1       
Th-230...........                     2                   54.1                2 x 10-4           5.41 x 10-3        7.6 x 10-4         2.1 x 10-2       
Th-231...........                     40                  1080                0.9                24.3               2.0 x 104          5.3 x 105        
Th-232...........                     Unlimited           Unlimited           Unlimited          Unlimited          4.0 x 10-9         1.1 x 10-7       
Th-234...........                     0.2                 5.41                0.2                5.41               8.6 x 102          2.3 x 104        
Th (natural).....                     Unlimited           Unlimited           Unlimited          Unlimited          8.1 x 10-9         2.2 x 10-7       
Ti-44............  Titanium(22).....  0.5                 13.5                0.2                5.41               6.4                1.7 x 102        
Tl-200...........  Thallium(81.1)...  0.8                 21.6                0.8                21.6               2.2 x 104          6.0 x 105        
Tl-201...........                     10                  270                 10                 270                7.9 x 103          2.1 x 105        
Tl-202...........                     2                   54.1                2                  54.1               2.0 x 103          5.3 x 104        
Tl-204...........                     4                   108                 0.5                13.5               1.7 x 101          4.6 x 102        
Tm-167...........  Thulium(69)......  7                   189                 7                  189                3.1 x 103          8.5 x 104        
Tm-168...........                     0.8                 21.6                0.8                21.6               3.1 x 102          8.3 x 103        
Tm-170...........                     4                   108                 0.5                13.5               2.2 x 102          6.0 x 103        
Tm-171...........                     40                  1080                10                 270                4.0 x 101          1.1 x 103        
U-230............  Uranium(92)......  40                  1080                1 x 10-2           0.270              1.0 x 103          2.7 x 104        
U-232............                     3                   81.1                3 x 10-4           8.11 x 10-3        8.3 x 10-1         2.2 x 101        
U-233............                     10                  270                 1 x 10-3           2.70 x 10-2        3.6 x 10-4         9.7 x 10-3       
U-234............                     10                  270                 1 x 10-3           2.70 x 10-2        2.3 x 10-4         6.2 x 10-3       
U-235............                     Unlimited           Unlimited           Unlimited          Unlimited          8.0 x 10-8         2.2 x 10-6       
U-236............                     10                  270                 1 x 10-3           2.70 x 10-2        2.4 x 10-6         6.5 x 10-5       
U-238............                     Unlimited           Unlimited           Unlimited          Unlimited          1.2 x 10-8         3.4 x 10-7       
U (natural)......                     Unlimited           Unlimited           Unlimited          Unlimited          2.6 x 10-8         7.1 x 10-7       
U (enriched 5% or                     Unlimited           Unlimited           Unlimited          Unlimited                             (See Table A-3)  
 less).                                                                                                                                                 
U (enriched more                      10                  270                 1 x 10-3           2.70 x 10-2                           (See Table A-3)  
 than 5%).                                                                                                                                              
U (depleted).....                     Unlimited           Unlimited           Unlimited          Unlimited                             (See Table A-3)  
V-48.............  Vanadium(23).....  0.3                 8.11                0.3                8.11               6.3 x 103          1.7 x 105        
V-49.............                     40                  1080                40                 1080               3.0 x 102          8.1 x 103        
W-178............  Tungsten(74).....  1                   27.0                1                  27.0               1.3 x 103          3.4 x 104        
W-181............                     30                  811                 30                 811                2.2 x 102          6.0 x 103        
W-185............                     40                  1080                0.9                24.3               3.5 x 102          9.4 x 103        
W-187............                     2                   54.1                0.5                13.5               2.6 x 104          7.0 x 105        
W-188............                     0.2                 5.41                0.2                5.41               3.7 x 102          1.0 x 104        
Xe-122...........  Xenon(54)........  0.2                 5.41                0.2                5.41               4.8 x 104          1.3 x 106        
Xe-123...........                     0.2                 5.41                0.2                5.41               4.4 x 105          1.2 x 107        
Xe-127...........                     4                   108                 4                  108                1.0 x 103          2.8 x 104        
Xe-131m..........                     40                  1080                40                 1080               3.1 x 103          8.4 x 104        
Xe-133...........                     20                  541                 20                 541                6.9 x 103          1.9 x 105        
Xe-135...........                     4                   108                 4                  108                9.5 x 104          2.6 x 106        
Y-87.............  Yttrium(39)......  2                   54.1                2                  54.1               1.7 x 104          4.5 x 105        
Y-88.............                     0.4                 10.8                0.4                10.8               5.2 x 102          1.4 x 104        
Y-90.............                     0.2                 5.41                0.2                5.41               2.0 x 104          5.4 x 105        
Y-91m............                     2                   54.1                2                  54.1               1.5 x 106          4.2 x 107        
Y-91.............                     0.3                 8.11                0.3                8.11               9.1 x 102          2.5 x 104        
Y-92.............                     0.2                 5.41                0.2                5.41               3.6 x 105          9.6 x 106        
Y-93.............                     0.2                 5.41                0.2                5.41               1.2 x 105          3.3 x 106        
Yb-169...........  Ytterbium(70)....  3                   81.1                3                  81.1               8.9 x 102          2.4 x 104        

[[Page 50289]]
                                                                                                                                                        
Yb-175...........                     30                  811                 0.9                24.3               6.6 x 103          1.8 x 105        
Zn-65............  Zinc(30).........  2                   54.1                2                  54.1               3.0 x 102          8.2 x 103        
Zn-69m...........                     2                   54.1                0.5                13.5               1.2 x 105          3.3 x 106        
Zn-69............                     4                   108                 0.5                13.5               1.8 x 106          4.9 x 107        
Zr-88............  Zirconium(40)....  3                   81.1                3                  81.1               6.6 x 102          1.8 x 104        
Zr-93............                     40                  1080                0.2                5.41               9.3 x 10-5         2.5 x 10-3       
Zr-95............                     1                   27.0                0.9                24.3               7.9 x 102          2.1 x 104        
Zr-97............                     0.3                 8.11                0.3                8.11               7.1 x 104          1.9 x 106        
--------------------------------------------------------------------------------------------------------------------------------------------------------
a International shipments of Einsteinium require multilateral approval of A1 and A2 values.                                                             
b International shipments of Fermium require multilateral approval of A1 and A2 values.                                                                 
c 20 Ci for Mo99 for domestic use.                                                                                                                      



                                    Table A-2.--General Values for A1 and A2                                    
----------------------------------------------------------------------------------------------------------------
                                                                  A1                            AA2             
                      Contents                       -----------------------------------------------------------
                                                          (TBq)         (Ci)          (TBq)            (Ci)     
----------------------------------------------------------------------------------------------------------------
Only beta- or gamma-emitting nuclides are known to            0.2           5     0.02           0.5            
 be present.                                                                                                    
Alpha-emitting nuclides are known to be present, or           0.10          2.70  2 x  x 10-5    5.41 x  x 10-4 
 no relevant data are available.                                                                                
----------------------------------------------------------------------------------------------------------------


           Table A-3.--Activity-mass Relationships for Uranium          
------------------------------------------------------------------------
                                           Specific activity            
Uranium enrichment \1\ wt % U-------------------------------------------
         235 present                      TBq/g                 Ci/g    
------------------------------------------------------------------------
0.45.........................  1.8 x  x 10-8               5.0 x  x 10-7
0.72.........................  2.6 x  x 10-8               7.1 x  x 10-7
1.0..........................  2.8 x  x 10-8               7.6 x  x 10-7
1.5..........................  3.7 x  x 10-8               1.0 x  x 10-6
5.0..........................  1.0 x  x 10-7               2.7 x  x 10-6
10.0.........................  1.8 x  x 10-7               4.8 x  x 10-6
20.0.........................  3.7 x  x 10-7               1.0 x  x 10-5
35.0.........................  7.4 x  x 10-7               2.0 x  x 10-5
50.0.........................  9.3 x  x 10-7               2.5 x  x 10-5
90.0.........................  2.2 x  x 10-6               5.8 x  x 10-5
93.0.........................  2.6 x  x 10-6               7.0 x  x 10-5
95.0.........................  3.4 x  x 10-6               9.1 x  x 10-5
------------------------------------------------------------------------
\1\ The figures for uranium include representative values for the       
  activity of the uranium-235 which is concentrated during the          
  enrichment process.                                                   

    Dated at Rockville, MD this 13th day of September 1995.

    For the Nuclear Regulatory Commission.
James M. Taylor,
Executive Director for Operations.
[FR Doc. 95-23538 Filed 9-27-95; 8:45 am]
BILLING CODE 7590-01-P