[7590-01-P]
NUCLEAR REGULATORY COMMISSION
10 CFR Part 71
[NRC-2016-0179]
RIN 3150-AJ85
Harmonization of Transportation Safety Requirements with IAEA Standards
AGENCY: Nuclear Regulatory Commission.

ACTION: Proposed rule and guidance; request for comment.

SUMMARY: The U.S. Nuclear Regulatory Commission (NRC), in consultation with the
U.S. Department of Transportation, is proposing to amend its regulations for the
packaging and transportation of radioactive material. The NRC has historically revised
its transportation safety regulations to ensure harmonization with the International
Atomic Energy Agency standards. These changes are necessary to maintain a
consistent regulatory framework with the U.S. Department of Transportation for the
domestic packaging and transportation of radioactive material and to ensure general
accord with International Atomic Energy Agency standards. Concurrently, the NRC is
issuing for public comment Draft Regulatory Guide DG-7011, which would become
Revision 3 to Regulatory Guide 7.9, “Standard Format and Content of Part 71
Applications for Approval of Packages for Radioactive Material.”

DATES: Submit comments by [INSERT DATE 75 DAYS AFTER DATE OF
PUBLICATION IN THE FEDERAL REGISTER]. Comments received after this date will
be considered if it is practical to do so, but the NRC is able to ensure consideration only
for comments received on or before this date.

ADDRESSES: You may submit comments by any of the following methods:
ï‚·

Federal rulemaking website: Go to https://www.regulations.gov and search

for Docket ID NRC-2016-0179. Address questions about NRC dockets to Dawn Forder;
telephone: 301-415-3407; email: Dawn.Forder@nrc.gov. For technical questions
contact the individual or individuals listed in the FOR FURTHER INFORMATION
CONTACT section of this document.
ï‚·

Email comments to: Rulemaking.Comments@nrc.gov. If you do not

receive an automatic email reply confirming receipt, then contact us at 301-415-1677.
ï‚·

Mail comments to: Secretary, U.S. Nuclear Regulatory Commission,

Washington, DC 20555-0001, ATTN: Rulemakings and Adjudications Staff.
For additional direction on obtaining information and submitting comments, see
“Obtaining Information and Submitting Comments” in the SUPPLEMENTARY
INFORMATION section of this document.

FOR FURTHER INFORMATION CONTACT: James Firth, 301-415-6628, email:
James.Firth@nrc.gov; or Bernard White, 301-415-6577, email: Bernard.White@nrc.gov.
Both are staff of the Office of Nuclear Material Safety and Safeguards, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001.

SUPPLEMENTARY INFORMATION:
TABLE OF CONTENTS:
I.
II.
III.

IV.
V.
VI.
VII.
VIII.
IX.

Obtaining Information and Submitting Comments
A. Obtaining Information
B. Submitting Comments
Background
Discussion
A. Action the NRC is Proposing to Take
B. Applicability of the Proposed Action
C. Discussion of Issues Specific Request for Comment
Specific Request for Comment
Section-by-Section Analysis
Regulatory Flexibility Certification
Regulatory Analysis
Backfitting and Issue Finality
Cumulative Effects of Regulation

X.
XI.

Plain Writing
Environmental Assessment and Proposed Finding of No Significant
Environmental Impact
XII.
Paperwork Reduction Act
XIII. Criminal Penalties
XIV. Coordination with NRC Agreement States
XV.
Compatibility of Agreement State Regulations
XVI. Voluntary Consensus Standards
XVII. Availability of Guidance
XVIII. Public Meeting
XIX. Availability of Documents

I.

Obtaining Information and Submitting Comments

A. Obtaining Information
Please refer to Docket ID NRC-2016-0179 when contacting the NRC about the
availability of information for this action. You may obtain publicly-available information
related to this action by any of the following methods:
ï‚·

Federal Rulemaking Website: Go to https://www.regulations.gov and

search for Docket ID NRC-2016-0179.
ï‚·

NRC’s Agencywide Documents Access and Management System

(ADAMS): You may obtain publicly-available documents online in the ADAMS Public
Documents collection at https://www.nrc.gov/reading-rm/adams.html. To begin the
search, select “Begin Web-based ADAMS Search.” For problems with ADAMS, please
contact the NRC’s Public Document Room (PDR) reference staff at 1-800-397-4209,
301-415-4737, or by email to PDR.Resource@nrc.gov. For the convenience of the
reader, instructions about obtaining materials referenced in this document are provided
in the “Availability of Documents” section.
ï‚·

NRC’s PDR: You may examine and purchase copies of public documents,

by appointment, at the PDR, Room P1 B35, One White Flint North, 11555 Rockville
Pike, Rockville, Maryland 20852. To make an appointment to visit the PDR, please send
an email to PDR.Resource@nrc.gov or call 1-800-397-4209 or 301-415-4737, between
8:00 a.m. and 4:00 p.m. eastern time (ET), Monday through Friday, except Federal

holidays.
B. Submitting Comments
Please include Docket ID NRC-2016-0179 in your comment submission.
The NRC cautions you not to include identifying or contact information that you
do not want to be publicly disclosed in your comment submission. The NRC will post all
comment submissions at https://www.regulations.gov as well as enter the comment
submissions into ADAMS. The NRC does not routinely edit comment submissions to
remove identifying or contact information.
If you are requesting or aggregating comments from other persons for
submission to the NRC, then you should inform those persons not to include identifying
or contact information that they do not want to be publicly disclosed in their comment
submission. Your request should state that the NRC does not routinely edit comment
submissions to remove such information before making the comment submissions
available to the public or entering the comment into ADAMS.

II.

Background

On June 12, 2015, the NRC, in consultation with the U.S. Department of
Transportation (DOT), published a final rule that amended the NRC’s regulations for the
packaging and transportation of radioactive material (80 FR 33988; June 12, 2015).
These amendments made conforming changes to the NRC’s regulations based on the
standards of the International Atomic Energy Agency (IAEA). That final rule, in
combination with a DOT final rule (79 FR 40589; July 11, 2014) amending title 49 of the
Code of Federal Regulations (49 CFR), brought U.S. regulations into general accord
with the 2009 Edition of the IAEA’s “Regulations for the Safe Transport of Radioactive
Material” (TS-R-1). The IAEA has since updated its standards for the transport of
radioactive material in “Regulations for the Safe Transport of Radioactive Material,”
Specific Safety Requirements No. 6 (SSR-6) (2012 and 2018 Editions).

The IAEA develops international safety standards for the safe transport of
radioactive material. The IAEA safety standards are developed in consultation with the
competent authorities of Member States, so they reflect an international consensus on
what is needed to provide for a high level of safety. By providing a global framework for
the consistent regulation of the transport of radioactive material, IAEA safety standards
facilitate international commerce and contribute to the safe conduct of international trade
involving radioactive material. By periodically revising its regulations to be compatible
with IAEA standards and DOT regulations, the NRC can remove inconsistencies that
could impede international commerce.
The roles of the DOT and the NRC in the coregulation of the transportation of
radioactive materials are documented in a Memorandum of Understanding (44 FR
38690; July 2, 1979). Because of the coregulation of the transportation of radioactive
materials in the United States, the NRC and the DOT have historically coordinated to
harmonize their respective regulations with the IAEA revisions through the rulemaking
process. In the NRC’s previous 10 CFR part 71 harmonization rulemaking, published in
the Federal Register on June 12, 2015, the Commission stated that the NRC will
consider any necessary changes related to SSR-6 in a future rulemaking after consulting
with DOT.
The NRC engaged with the DOT in the development of this proposed rule to
identify and evaluate gaps between 10 CFR part 71 regulations and the updated IAEA
standards in SSR-6, 2018 Edition. This proposed rule would close those gaps where
warranted. Harmonizing NRC regulations with the 2018 Edition of SSR-6 includes
changes made in the 2012 Edition of SSR-6 that have been carried forward to the 2018
Edition. The DOT is undertaking a similar initiative to harmonize its regulations in 49
CFR parts 107 and 171–180 with the 2018 Edition of SSR-6.
The NRC reviewed the 2018 Edition of SSR-6 and identified 10 regulatory issues
for harmonization with the IAEA and another 4 NRC-initiated changes to 10 CFR part 71
to be evaluated during the rulemaking development process. Fourteen of these issues

were documented in the “Issues Paper on Potential Revisions to Transportation Safety
Requirements and Harmonization with International Atomic Energy Agency
Transportation Requirements” (issues paper). The issues paper, public meeting, and
request for comment were published in the Federal Register (81 FR 83171; November
21, 2016). The NRC held a public meeting on December 5–6, 2016, to discuss the
issues paper, and the DOT participated in that public meeting. A summary of the public
meeting, including the attendance list, was issued on December 14, 2016. After the
public meeting, the NRC received 49 comment submissions on the issues paper
identified comments that are pertinent to this proposed rule, and considered these
comments in the development of a draft regulatory basis. In addition to the 14 issues
documented in the paper, the NRC identified other potential changes to the regulations,
including clarifications to ensure compatibility with the DOT and changes to the
compatibility categories for Agreement State regulations. These potential changes were
grouped under a new issue that was designated as Issue 15 in the draft regulatory basis.
All 15 issues are described in Section III of this document.
On April 12, 2019, the NRC published the draft regulatory basis for this proposed
rule in the Federal Register and requested public comments (84 FR 14898; April 12,
2019). In the regulatory basis, the NRC evaluated four alternative actions for each
issue. These were: Alternative 1–take no action and maintain the status quo;
Alternative 2–issue generic communications and regulatory guidance; Alternative 3–
issue license-specific conditions and exemptions; and Alternative 4–initiate a rulemaking
action to revise 10 CFR part 71. The alternatives were evaluated based on their viability
to resolve the regulatory issues of concern and estimates of their costs and potential
benefits. The NRC determined that the rulemaking action, Alternative 4, for Issues 1 (in
part), 2, and 4–15, in combination with the no-action alternative, Alternative 1, for Issue
3, was the NRC-recommended action because it represented the most effective and
least-costly option. Alternatives 2 and 3 would not address all of the regulatory issues or
would result in higher costs to the NRC and industry.

The NRC also held a public meeting on April 30, 2019, to discuss the draft
regulatory basis and answer questions. The NRC received seven public comment
submissions on the draft regulatory basis–three with general comments on the
rulemaking and four with comments on specific issues–as well as comments that were
considered outside the scope of this proposed rule. All three general comments were
supportive of the harmonization effort with IAEA SSR-6. The NRC did not receive any
comments on Issues 2, 6, and 14. The NRC received comments supportive of the
proposal for Issues 4b, 11, 12, 13 and 15, along with comments supportive of other
issues which also recommended modifications to the NRC’s proposed changes. One
comment on Issue 5 proposed the NRC add a definition of “radiation level” to 10 CFR
part 71, which the NRC included in this proposed rule.
One comment on Issue 1 stated that the fissile exemption mass limits in 10 CFR
part 71 should match those in SSR-6, paragraph 417, to avoid confusion for international
shipments from the United States. The NRC has determined that its regulations for
fissile exemption mass limits should differ from the IAEA’s requirements to provide
flexibility for shippers. Specifically, the NRC requirements in this proposed rule would
adopt a 3.5-gram limit from SSR-6, paragraph 417(c), but without the associated
consignment limit found in paragraph 570(c); they also would adopt a higher mass limit
than SSR-6, paragraph 417(e). Several existing fissile exemptions under § 71.15 do not
have corresponding exceptions under SSR-6, paragraph 417; if the NRC made 10 CFR
part 71 fissile exemptions identical to the fissile exceptions in SSR-6, paragraph 417,
fissile material licensees would lose the benefit of these exemptions. Also, the NRC is
not pursuing the competent authority-approved exception in SSR-6, paragraph 417(f).
The NRC has determined that the current fissile exemptions under § 71.15 provide
flexibility for shipping low masses or concentrations of fissile materials, and licensees
can submit a specific exemption request under § 71.12 for fissile materials that do not
meet the fissile exemption criteria in § 71.15.
The NRC received comments on Issues 4 and 8 which suggested that the NRC

“grandfather” packages from having to meet the revised requirements. The NRC is
proposing to “grandfather” older packages as discussed in Issue 10, “Transitional
Arrangements.”
Comments on Issue 4 on the proposed insolation requirements stated that these
requirements would present challenges to certificate holders, including cost to
certificate holders to evaluate the new conditions; changing the units without revising
the corresponding values may result in decreasing margins or exceeding thermal limits;
and the insolation values are referenced in other documents, which may have an
impact to the thermal evaluations for storage systems certified under 10 CFR part 72.
While the NRC agrees there will be costs with evaluating the new insolation
requirements, the NRC estimates that the cost for existing certificates to show
compliance with the revised insolation will be small, since the increased insolation load
would be approximately 3 percent. In addition, harmonizing NRC requirements with
those of IAEA will ensure that packages approved by the NRC would also be
acceptable in other countries where they might be used for international transport. The
NRC made no changes as a result of this comment. The NRC recognizes that all
packages age over time and that aging effects should be considered for all packages,
not just for dual-purpose packages.
The NRC received comments on Issue 9 opposing the addition of an aging
management program to 10 CFR part 71. The commenters stated that, if such a
program were added, the program should be limited to packages other than dualpurpose spent nuclear fuel packages/canisters. The NRC is not proposing to impose a
requirement for an aging management plan. The proposed rule includes requirements
that aging effects are evaluated in the application for approval and that the application
for approval include a maintenance program. Another comment on Issue 9 supported
evaluating aging effects but only for dual-purpose spent fuel packages, excluding
packages that are not kept in long-term storage prior to transport.
One comment on Issue 10 supported phasing out older packages as proposed

in transitional arrangements but suggested a phase-out period longer than 4 years.
The NRC agreed and is proposing an 8-year phase out of older packages. As part of
the NRC’s 2004 amendment to 10 CFR part 71 (69 FR 3697; January 26, 2004),
certain transportation packages, those compatible with the 1967 edition of Safety
Series No. 6, became unauthorized for use under the 10 CFR part 71 general license
after October 1, 2008. The NRC received requests to extend the phase-out date
beyond the initial 4-year period to allow sufficient time to design, obtain approval for,
and fabricate new packages. Given this experience, in this proposed rule, the NRC has
selected a phase-out period of 8 years to give certificate holders sufficient time to
conduct these activities, if needed. The NRC estimates that it could take 2 to 4 years
for design of a new package and preparation of an application, 1 to 2 years for package
approval, and 1 to 2 years for package fabrication, depending on the package’s
complexity. Another comment on Issue 10 on transitional arrangements stated that the
NRC should not phase out packages with a “-96” in the package identification number
and that the proposed phase out of packages did not consider the cost impact for
designing new packages. The NRC is not proposing to phase out packages with a
“-96” in the proposed rule, but rather proposing to phase out packages that do not have
either a “-85” or a “-96” in the package identification number (i.e., packages approved
before April 1, 1996). The NRC included the cost of designing a new package in the
regulatory analysis for the proposed rule.
The NRC received one comment on Issue 12 on the proposed quality
assurance program (QAP) changes, stating that the proposed change would be
duplicative with 10 CFR part 50 QAP requirements. The NRC disagrees with this
comment because if a 10 CFR part 50 licensee uses its 10 CFR part 50 QAP for 10
CFR part 71 activities, the QAP reporting requirements in 10 CFR part 50 would be
controlling and 10 CFR part 71 QAP reporting requirements would not apply. Also, the
NRC notes that many users of 10 CFR part 71 do not have 10 CFR part 50 licenses,
and the 10 CFR part 71 QAP change provisions would not be duplicative for them.

The NRC received a comment on Issue 15 on the advance notification
requirements in § 71.97, stating that there is no actual provision requiring advance
notification for spent fuel shipments. The requirements in § 71.97 currently contain
reporting requirements that are duplicative with those in 10 CFR part 73, and the NRC is
proposing to delete the duplicative language.
Because none of the comments would result in significant changes to the draft
regulatory basis, the NRC considered these comments in preparing this proposed rule
and did not issue a final regulatory basis.

III.

A.

Discussion

Action the NRC is Proposing to Take
The NRC is proposing to amend its regulations to harmonize them with the IAEA

international transportation standard No. SSR-6 (2018 Edition). These revisions would
be coordinated with DOT and its hazardous materials regulations to maintain a
consistent framework for the domestic transportation and packaging of radioactive
material.
This proposed rule also would revise 10 CFR part 71 to include administrative,
editorial, or clarifying changes, including changes to certain Agreement State
compatibility category designations that are further discussed in Section XV,
“Compatibility of Agreement State Regulations,” of this document.
B.

Applicability of the Proposed Action
This action would affect 1) NRC licensees authorized by a Commission-issued

specific or general license to receive, possess, use, or transfer licensed material, if the
licensee delivers that material to a carrier for transport, or transports the material outside
of the site of usage as specified in the NRC license, or transports that material on public
highways; 2) holders of, and applicants for, a certificate of compliance (CoC) under 10
CFR part 71; and 3) holders of a 10 CFR part 71 QAP approval. This action also would
change requirements that are a matter of compatibility with the Agreement States.

Therefore, the Agreement States would need to update their regulations, as appropriate,
at which time those licensees in Agreement States would need to meet the compatible
Agreement State regulations.
C.

Discussion of Issues
The NRC is proposing to revise 10 CFR part 71 as described in the 15 issues

listed in this document and summarized in the following table (note that the issue
numbers described in Section III.C of this document are consistent with those described
in the regulatory basis):
Issue
IAEA
Harmonization
X

DOT
Harmonization

Other
Changes

No
Action

X

X

4.1

X

4.2

X

X

X

X

X

X

X

X

X

X

X

X

X

X

X

15.1

X

15.2

X

15.3
15.4

X

X
X

15.5

X

Issue 1. Revision of Fissile Exemptions
The fissile material exemptions in § 71.15 and the fissile material general
licenses in §§ 71.22 and 71.23 allow licensees to ship low-risk fissile material (e.g., small
quantities or low concentrations) without meeting the fissile material packaging
requirements and criticality safety assessments, as specified in §§ 71.55 and 71.59, and
without obtaining prior NRC approval. For these low-risk fissile material shipments, the
fissile material exemptions and general licenses provide reasonable assurance that
criticality safety is afforded under normal conditions of transport and hypothetical
accident conditions. In 2012, IAEA modified the fissile exception provisions in SSR-6,
paragraph 417, to include three new per-package mass limit options, with associated
mass limits on the consignment and/or conveyance.
The NRC proposes to incorporate two additional fissile exemptions under
§ 71.15. This proposed rule would adopt the exception in SSR-6, paragraph 417(c),
without the associated consignment limit of IAEA SSR-6, paragraph 570(c). This
proposed rule would also adopt the exception in SSR-6, paragraph 417(e), with its
associated exclusive use restriction in paragraph 570(e), but with a higher mass limit.
Since the amount of fissile material allowed by SSR-6, paragraph 417(c), is
similar to the existing exemption in § 71.15(a), in terms of reactivity, the NRC determined
that the consignment limit of IAEA SSR-6, paragraph 570(c), is not necessary.
Consignment limits, as provided in 570(c), do not prevent the accumulation of packages
on a transport conveyance, as there is no limit to the number of consignments that may
be present on a single conveyance. Additionally, the number of these packages does
not need to be limited by regulation because reaching the amount required to approach
criticality on a single conveyance is not credible.
The NRC has determined that a mass value higher than that contained in IAEA
SSR-6, paragraph 417(e), is justified, given the conservatism inherent in the exclusive

use restriction of the SSR-6 provision, and in basing the mass limit on plutonium-239
(239Pu), which would have to be shipped in a Type B package. The NRC proposes a
limit of 140 grams of fissile material on a conveyance shipped under exclusive use, as
another exemption under § 71.15. This limit is based on one fifth of a minimum critical
mass of uranium-235 (235U) (as defined in American National Standards Institute/
American Nuclear Society [ANSI/ANS] 8.1-2014 (Reaffirmed 2018), “Nuclear Criticality
Safety in Operations with Fissionable Materials Outside Reactors”) under optimum
conditions. This mass represents a conservative limit for fissile material, since five times
this amount would remain subcritical under any condition. Additionally, the limit provides
safety equivalent to packages approved under 10 CFR part 71 and could provide more
flexibility for shipping individual contaminated items or small quantities of fissile material.
The NRC considers 235U for this limit rather than 239Pu, as any amount of 239Pu over
0.435 grams is considered Type B, which would have to be packaged to withstand both
normal and hypothetical accident conditions of transport. Although the NRC proposed
value is different from the IAEA SSR-6, paragraph 417(e), value, the NRC determined
that the higher value is technically justified and will be appropriate for NRC licensees
who ship specific waste streams (e.g., decommissioning waste), and that there will be
little international shipment from the United States of this type of material. Licensees
who ship material internationally must comply with DOT requirements for the use of
international standards in title 49, “Transportation,” of the CFR.
Additionally, the NRC is not proposing to adopt the “packaged or unpackaged”
language in the fissile exception provision of IAEA SSR-6, paragraph 417(e). The 140gram limit, as with other fissile exemption provisions in § 71.15, only relieves the
consignor from having to ship in a “Fissile” package, evaluated per the requirements of
§§ 71.55 and 71.59. This material is still subject to all other radioactive materials
transportation requirements in 10 CFR part 71 and in 49 CFR part 173 and should be
packaged accordingly. The NRC is proposing to make a minor change to § 71.15(d) for
clarity and to maintain consistent language throughout § 71.15.

Issue 2. Revision of Reduced External Pressure Test for Normal Conditions of
Transport
The regulation at § 71.71(c)(3) requires Type AF and Type B package designs to
be able to withstand a reduction in external pressure to 25 kilopascals (kPa) (3.6 psia)
under normal conditions of transport. For a Type A package (as defined in SSR-6,
paragraphs 231 and 429; 10 CFR 71.4, “Definitions”; or 49 CFR 173.403, “Definitions”),
IAEA SSR-6, paragraph 645, states that “[t]he containment system shall retain its
radioactive contents under a reduction of ambient pressure to 60 kPa.” This requirement
also applies to Type B(U) and Type B(M) packages, in accordance with SSR-6,
paragraphs 652 and 667, respectively. Additionally, IAEA SSR-6, paragraph 621,
indicates packages containing radioactive material to be transported by air shall be
capable of withstanding, without loss or dispersal of the radioactive contents from the
containment system, an internal pressure that produces a pressure differential of not
less than maximum normal operating pressure plus 95 kPa (13.8 psi).
In a final rule published by the DOT (79 FR 40589; July 11, 2014), the DOT
harmonized its regulations in 49 CFR chapter I to the 2009 Edition of IAEA TS-R-1. In
that final rule, the DOT explained that a Type A package must be designed to ensure the
package can retain its contents under the reduction of ambient pressure. That ambient
pressure value, found at 49 CFR 173.412(f), was changed from 25 kPa (3.6 psia) to 60
kPa (8.7 psia).
The NRC considered whether it should change the reduced external pressure
test requirement in § 71.71(c)(3) to harmonize with the IAEA transport standards and to
be consistent with the DOT regulations for design requirements for Type A packages.
The NRC assessed the potential impacts of the change in the external pressure value
from 25 kPa (3.6 psia) to 60 kPa (8.7 psia) and the additional air transport requirements
from SSR-6, paragraph 621. The current NRC reduced external pressure test
requirement, 25 kPa (3.6 psia), equates to an altitude of about 35,000 feet (10,668
meters) above sea level, which is an appropriate altitude for air transport of packages.

Since cargo planes use pressurized cargo holds during air transport, this external
pressure value also represents the ambient pressure on a package should the cargo
hold depressurize. Whereas the 60 kPa (8.7 psia) value equates to an altitude of about
14,040 feet (4,279 meters) above sea level. Thus, while the 60 kPa (8.7 psia) external
pressure value equates well with the highest paved road in the United States (14,130
feet (4,307 meters)) and with the elevation of the highest operating freight railroad in the
United States (La Veta Pass at 9,242 feet (2,817 meters)), it would not support air
transport conditions, as cargo planes operate at higher altitudes. When comparing the
current 25 kPa (3.6 psia) value with the proposed 60 kPa (8.7 psia) value, and the
associated altitudes, the NRC determined that no change to § 71.71(c)(3) is needed,
and the 25 kPa (3.6 psia) value should be retained.
The NRC also considered adding the air transport requirements from SSR-6,
paragraph 621. However, other than specific air transport requirements at § 71.55(f),
“General requirements for fissile material packages” and § 71.88, “Air transport of
plutonium,” 10 CFR part 71 does not contain “mode-specific” regulations. Because the
existing reduced external pressure test value covers air transport conditions as
discussed above, and because of the robustness of Type AF and Type B packages, as
compared to Type A packages, the NRC finds it unnecessary to add the mode-specific
air transport requirements from SSR-6, paragraph 621, into 10 CFR part 71.
Based on the above considerations and assessments, the NRC has decided not
to pursue any changes to § 71.71(c)(3). As a result, no further discussion or analysis is
presented in this proposed rule on the reduced external pressure test for normal
conditions of transport.
Issue 3. Inclusion of Type C Package Standards
In the 2004 final rule, the NRC did not adopt the regulations for Type C packages
contained in IAEA TS-R-1. The NRC did not adopt them because 1) §§ 71.64 and 71.74
for plutonium air transportation contain more rigorous packaging standards, 2) the NRC
perceived no need (current or anticipated) for such packages, and 3) if a need arose for

import or export, it could be accomplished through the DOT regulations.
In the request for comment on the issues paper, the NRC asked stakeholders
whether there was a need for domestic transport of Type C packages. No NRC
licensees expressed a need for domestic transport of Type C packages. Therefore, the
NRC has decided not to pursue further changes to Type C package standards as
contemplated in the regulatory basis document. As a result, no further discussion or
analysis is presented in this proposed rule on that issue.
Issue 4. Revision of Insolation Requirements for Package Evaluations
During transport, a package is subjected to heating by the sun, called insolation.
The effect of insolation is an increase in the package temperature. The NRC is
proposing to change the unit of measure for the values of insolation used for the heat
test for normal conditions of transport in § 71.71(c)(1), and to add insolation to the initial
conditions for the tests for hypothetical accident conditions in § 71.73(b).
Issue 4.1. Revision of Units for Insolation for Normal Conditions of Transport
The units for insolation in 10 CFR part 71 are gram calories per square
centimeter (g cal/cm2). When the IAEA published Safety Series No. 6, “Regulations for
the Safe Transport of Radioactive Material, 1985 Edition,” it revised the units used for
insolation for normal conditions of transport from a hybrid of English and metric units
(g cal/cm2) to metric units (watts per square meter (W/m2)). When the IAEA changed the
units, it chose to keep the same numerical values, thus increasing the evaluated solar
heat load on a package by approximately 3 percent. The IAEA did not provide a
technical rationale for this change; however, the NRC observes that retaining the
existing numerical quantities maintains simple (round) values in the regulations that
result in a small change in solar heat load.
The NRC previously harmonized its regulations with the 1985 Edition of Safety
Series No. 6 (60 FR 50248; September 28, 1995). That final rule neither discussed nor
proposed changing the units on the heat test for normal conditions of transport in
§ 71.71(c)(1). Consequently, the current units for insolation in 10 CFR part 71 are

“g cal/cm2.” This is inconsistent with IAEA standards in the 2018 Edition of SSR-6. As a
result, NRC package approvals are evaluated for less insolation than that prescribed by
IAEA standards and evaluated for approval by foreign competent authorities.
The NRC is proposing to revise the units of insolation for the heat test for normal
conditions of transport in § 71.71(c)(1) to match the units used in the 2018 Edition of
SSR-6 to ensure that NRC requirements for insolation are consistent with the IAEA
standard. Consistent with Issue 10, “Transitional Arrangements,” the NRC would not
expect a certificate holder to evaluate the higher solar heat load unless it requests a
revision of its certificate to show compliance with the revised transportation regulations
in 10 CFR part 71. Additionally, given the small increase in insolation due to the revised
units, the NRC expects that certificate holders will be able to show compliance with the
package approval standards in subpart E, “Package Approval Standards,” to 10 CFR
part 71.
Issue 4.2. Inclusion of Insolation for Hypothetical Accident Conditions
In Safety Series No. 6, “Regulations for the Safe Transport of Radioactive
Material, 1985 Edition (As Amended 1990),” paragraph 628 stated, “With respect to the
initial conditions for the thermal test, the demonstration of compliance shall be based
upon the assumption that the package is in equilibrium at an ambient temperature of 38
°C. The effects of solar radiation may be neglected prior to and during the tests, but
must be taken into account in the subsequent evaluation of the package response.”
The thermal test, previously in paragraph 628, was moved to paragraph 728 in
the 1996 Edition of TS-R-1 and revised to state, “The specimen shall be in thermal
equilibrium under conditions of an ambient temperature of 38 °C, subject to the solar
insolation conditions specified in Table XI and subject to the design maximum rate of
internal heat generation within the package from the radioactive contents.”
When the NRC revised its regulations in 2004 to harmonize with the 1996 IAEA
standards (69 FR 3697; January 26, 2004), the NRC did not revise the initial conditions
of the fire test listed in § 71.73(b) to require evaluation of insolation as an initial

condition.
Since a fire can occur on a hot, sunny day, and to be consistent with IAEA
standards, the NRC is proposing to revise the initial conditions in § 71.73(b) to require
insolation as an initial condition for all the tests for hypothetical accident conditions.
Consistent with Issue 10, “Transitional Arrangements,” the NRC would expect a
certificate holder to evaluate the revised initial conditions in § 71.73 if it wants to revise
its certificate to show compliance with the revised transportation regulations in 10 CFR
part 71.
Issue 5. Inclusion of Definition for Radiation Level
The term “radiation level” was first introduced in the IAEA transport standards in
Safety Series No. 6, 1973 Edition, and it was defined in terms of “dose-equivalent rate”
as “the corresponding radiation dose-equivalent rate expressed in millirem per hour.”
External radiation standards were defined in terms of radiation levels in each
subsequent edition of the IAEA’s transport standards, including the 2012 Edition of SSR6. In the 2018 Edition of SSR-6, the IAEA replaced the term “radiation level” with the
term “dose rate” and defined the dose rate to be the dose-equivalent per unit time.
Because the current regulations in 10 CFR part 71 use the term “radiation level,” the
NRC is concerned that using a different term from the IAEA to define external radiation
standards could create some confusion with respect to international shipments.
Additionally, NRC regulations in 10 CFR part 20, “Standards for Protection
Against Radiation,” include a definition for “dose equivalent” in § 20.1003 that means the
product of the absorbed dose in tissue, quality factor, and all other necessary modifying
factors at the location of interest. The units of dose equivalent are the rem and sievert
(Sv).
The NRC considered replacing the term “radiation level” used throughout 10 CFR
part 71 with “dose equivalent rate.” However, this change would result in cost impacts to
licensees to change documentation and training programs with no safety benefit.
Therefore, in order to minimize the burden to licensees, the NRC is proposing to add a

definition to § 71.4 that clarifies that “radiation level” means “dose equivalent rate,” which
enables the NRC to continue using “radiation level” throughout 10 CFR part 71. The
NRC is not expecting any licensee to change its documentation to account for this new
definition.
Issue 6. Deletion of Low Specific Activity-III Leaching Test
The definition for “Low Specific Activity (LSA) material” in § 71.4 includes three
categories of material: LSA-I, LSA-II, and LSA-III. Radioactive material, low specific
activity category III (i.e., LSA-III) includes solids, excluding powders, that meet the
requirements in § 71.77, “Qualification of LSA-III material” and which have an estimated
average specific activity limit that does not exceed 2 x 10-3 times the A2 value per gram
(A2/g). The qualification tests in § 71.77 include a leaching test with immersion of the
specimen material for 7 days. The IAEA eliminated the LSA-III leaching test in SSR-6,
2018 Edition, from paragraphs 409, 601, and 701. Consequently, the NRC is proposing
corresponding revisions to §§ 71.4, 71.77, and 71.100, “Criminal penalties,” to remove
the leaching test and its references.
In April 2015, an international working group meeting was conducted to discuss
issues related to LSA-II and LSA-III material, with special attention on the need for the
LSA-III leaching test. The need for the leaching test was questioned because the
working group determined that the test has no bearing on the inhalation risk of exposure
to material during transport. The inhalation risk is used to determine the average
specific activity limits for both LSA-II and LSA-III material, which are 10-4A2/g and 2 x 103A /g,
respectively. Related investigations dating back to 2003 revealed that the amount

of released radioactive material leading to an inhalation dose under the mechanical tests
for normal conditions of transport greatly depend on the physical form of the LSA
material. The primary difference between LSA-II and LSA-III materials is that LSA-III is
limited to solid material, excluding powders. Due to the solid nature of the LSA-III
material, the amount of airborne radioactivity released during the mechanical tests for
normal conditions of transport leading to an inhalation dose is at least a factor of 100

lower for LSA-III solids than for LSA-II solids in powder form. This much lower airborne
release for LSA-III material due to its non-readily dispersible form outweighs the
difference in average specific activity limit, which is 20 times greater for LSA-III
compared to LSA-II material in powder form. Because of the non-dispersible form of the
LSA-III material, the working group determined that there was no need to take credit
from a leaching test to justify this allowable 20-fold increase in average specific activity
between LSA-III and LSA-II material.
The NRC recognizes the working group’s information, and is recommending
harmonization with SSR-6, 2018 Edition, and removal of the leaching test from 10 CFR
part 71. The NRC agrees that requiring the LSA-III leaching test does not increase the
safety of the material during transport. Further, the test does not decrease the inhalation
pathway exposure when compared to LSA-II material in powder form, and therefore
should be removed from 10 CFR part 71. The NRC considered the information provided
by the LSA-II and LSA-III working groups and comments received on this issue during
the comment period on the NRC’s issues paper. Additionally, the NRC considers that
removal of the leaching test also would reduce regulatory burden for shippers, while still
maintaining reasonable assurance of safety for transport of LSA-III material.
The NRC is proposing to remove the leaching test in § 71.77 and make
conforming changes to §§ 71.4 and 71.100, which both reference § 71.77.
Issue 7. Inclusion of New Definition for Surface Contaminated Object
As more nuclear facilities begin decommissioning activities, there will be an
increase in the number of shipments of radioactive materials from these facilities.
Decommissioning activities can include transporting large radioactive objects (e.g.,
steam generators, coolant pumps, and pressurizers). Under current NRC regulations,
shipment of such large, nonstandard packages that do not meet the existing definition of
surface contaminated objects (i.e., either SCO-I or SCO-II, as defined in § 71.4) could be
addressed through a special package authorization under § 71.41(d). However, such an
authorization may take significant time. The NRC proposes to add a regulatory definition

for SCO-III to include these types of objects, allowing a shipper to more appropriately
categorize the item it is planning to transport. The NRC anticipates an increase in
efficiency for both the NRC and licensees when the SCO-III definition is included in 10
CFR part 71 when compared to the special package authorization review needed under
§ 71.41(d). Harmonization with SSR-6, 2018 Edition, would add the new SCO-III
category and the associated definition.
In the 2004 final rule (69 FR 3697; January 26, 2004), the NRC determined that
special package authorizations were necessary because there were no regulatory
provisions in 10 CFR part 71 concerning large, nonstandard packages considered for
transportation. Therefore, the NRC added paragraph (d) to § 71.41. Since that time, the
NRC has gained experience with the safety aspects of shipping these types of large,
non-standard packages. For example, in 2006, the LaCrosse reactor vessel was the
first shipment in which a package was approved under § 71.41(d). In addition, a special
package authorization was issued for the West Valley Melter Package from the West
Valley Demonstration Project. In the future, a licensee shipping large radioactive objects
that have been determined to meet the definition of SCO-III would not need NRC review
and approval for a special package authorization.
Both the NRC and DOT intend to add a definition for SCO-III. The NRC is
coordinating with the DOT to align its definition with the DOT’s, since the DOT is the lead
agency for review and evaluation of both LSA and SCO material.
Issue 8. Revision of Uranium Hexafluoride Package Requirements
In the 2004 final rule (69 FR 3697; January 26, 2004), the NRC harmonized its
regulations with the 1996 Edition of IAEA TS-R-1. In that final rule, the NRC added a
new provision, § 71.55(g), to provide a specific exception for certain uranium
hexafluoride (UF6) packages from the requirements of § 71.55(b). The exception allows
UF6 packages to be evaluated for criticality safety without considering inleakage of water
into the containment system, provided certain conditions are met, including that the
uranium is enriched to not more than 5 weight percent in 235U. To use this exception, the

applicant must demonstrate, among other things, that, following the tests for hypothetical
accident conditions in § 71.73, there is no physical contact between the valve body and
any other component of the packaging, other than at its original point of attachment, and
the valve remains leak tight. “Leaktight” is defined in ANSI N14.5-2014, “American
National Standard for Radioactive Materials — Leakage Tests on Packages for
Shipment,” as “[t]he degree of package containment that, in a practical sense, precludes
any significant release of radioactive materials. This degree of containment is achieved
by demonstration of a leakage rate less than or equal to 1×10-7 ref·cm3/s, of air at an
upstream pressure of 1 atmosphere (atm) absolute (abs), and a downstream pressure of
0.01 atm abs or less.”
The NRC provided the specific exception: 1) to be consistent with the worldwide
practice and limits established in national and international standards (ANSI N14.1-2012,
“Nuclear Materials - Uranium Hexafluoride — Packagings for Transport,” and
International Organization for Standardization 7195, “Packaging of Uranium Hexafluoride
(UF6) for Transport”) and DOT regulations (49 CFR 173.417(b)(5)); 2) because of the
history of safe shipment; and 3) because of the essential need to transport the
commodity. In that final rule, the NRC codified its long-standing practice to not consider
water inleakage into UF6 packages as long as the documentation of the results of the
tests for hypothetical accident conditions tests at § 71.73 show that the cylinder valve
was not affected.
In SSR-6, 2018 Edition, the IAEA added the same standard for the plug as was
added in the 1996 Edition of TS-R-1 for the valve to ensure that the entire cylinder
remains leak tight. The revised paragraph 680(b)(i), SSR-6, 2018 Edition, states:
“Packages where, following the tests prescribed in para. 685(b), there is no physical
contact between the valve or the plug and any other component of the packaging other
than at its original point of attachment and where, in addition, following the test
prescribed in para. 728, the valve and the plug remain leaktight.”
The 30-inch UF6 cylinder, the most commonly used cylinder to transport large

quantities of enriched UF6 for the fuel fabrication industry, has two penetrations: one for
the valve at the top to fill the cylinder and one for the drain plug at the bottom used
during maintenance. In order to ensure criticality safety, both the plug and the valve
must remain leak tight after the tests for hypothetical accident conditions to prevent
ingress of water into the cylinder. While this may be a new requirement in transportation
regulations, during package approval, the NRC has always verified that the entire 30B
cylinder remained leak tight after the tests for hypothetical accident conditions.
The NRC is proposing to revise § 71.55(g)(1) to require that there is no contact
between the cylinder plug and any other part of the packaging, other than at its original
attachment point and that the cylinder plug remains leak tight, as NRC requires for the
cylinder valve.
Issue 9. Inclusion of Evaluation of Aging Mechanisms and a Maintenance Program
The NRC regulations do not explicitly require that a package application include
an evaluation of aging mechanisms and a maintenance program. Rather, applicants
include an evaluation of aging effects on package components to ensure there is no
significant degradation in accordance with § 71.43(d). The NRC regulations at
§ 71.43(d) require that packages be made of materials and construction that assure that
there will be no significant chemical, galvanic, or other reaction (including effects of
irradiation from the package contents) 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.
For those components where aging is detrimental to package performance,
applicants provide a description of the maintenance program, including periodic testing
to evaluate the components’ efficacy and/or a replacement or repair schedule, to
mitigate those detrimental effects. The NRC requires that licensees and CoC holders
follow the maintenance program, which is provided in the application for approval, as a
condition of approval in the CoC. Additionally, NRC regulations at § 71.87(b) require

that, prior to each shipment, the licensee ensures that the package is in unimpaired
physical condition except for superficial defects such as marks or dents. Meeting this
regulation, along with the scheduled periodic tests and replacement/repair in the
maintenance program, should identify package deterioration prior to age-related
degradation becoming a safety issue during transport.
In paragraph 613A, SSR-6, 2018 Edition, the IAEA added that package design
evaluations must consider aging mechanisms. In paragraph 809, SSR-6, 2018 Edition,
the IAEA added that the application for package approval must contain a maintenance
program. Because an evaluation of aging effects and a description of the maintenance
program are not specifically required by 10 CFR part 71, the NRC is proposing to revise
§ 71.43(d) to specifically include the evaluation of the effects of aging, and add a new
provision to subpart D, “Application for Package Approval,” to include a description of the
maintenance program in an application for package approval, to better align with these
standards in SSR-6, 2018 Edition.
Issue 10. Revision of Transitional Arrangements
Historically, IAEA standards and DOT and NRC regulations have included
transitional arrangements when the regulations have undergone revision. The purpose
is to minimize the costs and impacts of implementing changes in the regulations, since
package designs and special form sources that are compliant with the existing
regulations do not become unsafe when the regulations are revised (unless a significant
safety issue is corrected in the revision).
Typically, the transitional arrangements include provisions that allow for
1) continued use of existing package designs and packagings already fabricated; and
completion of packagings in the process of being fabricated, although some restrictions
on fabrication of packagings approved to earlier editions of the regulations may be
imposed; 2) restriction on modifications to package designs without the need to
demonstrate full compliance with the revised regulations; 3) changes in packaging
identification numbers; and 4) changes to the fabrication and use of special form sources

approved to earlier versions of the regulations.
The NRC CoCs include a package identification number which identifies the NRC
regulations and the corresponding version of IAEA standards to which the package was
approved. For example, packages with a “-85” in the package identification number
were approved to NRC regulations compatible with the provisions of the 1985 or 1985
(as amended 1990) Editions of Safety Series No. 6. NRC packages with a “-96” in the
package identification number were approved to NRC regulations compatible with the
1996 Edition of TS-R-1.
The IAEA updated its transitional arrangements in paragraphs 819–823, SSR-6,
2018 Edition, for packages that have a “-85” or “-96” in their package identification
number. However, it does not include transitional arrangements for package designs
approved under the IAEA’s 1973 Edition of Safety Series No. 6, “Regulations for the
Safe Transport of Radioactive Materials.” The NRC previously harmonized its
requirements with the 1973 Edition; corresponding packages are those for which the
CoC does not have a year designation in the package identification number. By not
including transitional arrangements on these packages, the IAEA standards effectively
phase out the use of these packages approved under the 1973 Edition of Safety Series
No. 6.
The IAEA’s SSR-6, 2018 Edition, also prohibits, after December 31, 2028, the
fabrication of new packagings that have not been shown to meet SSR-6, 2018 Edition
standards. This means that package designs approved to earlier versions of IAEA
standards (i.e., NRC-approved packages for which the CoC has a “-96” in its package
identification number), could not be used unless fabrication is completed before January
1, 2029. Note that IAEA standards and NRC regulations already prohibit the use of
packages that have “-85” in their package identification number on the CoC if their
fabrication was not completed by December 31, 2006.
The IAEA’s SSR-6, 2018 Edition, also phases out certain special form
radioactive material. The NRC regulations contain a definition of, and the tests for,

special form radioactive material. Special form radioactive material is either a nondispersible solid or sealed in a capsule so that the dispersibility, and therefore the
radiological hazard, of the radioactive material is diminished. In order to be designated
as special form, the radioactive material must be evaluated using the tests and
acceptance criteria in § 71.75.
Paragraph 823 of SSR-6, 2018 Edition, does not include provisions for use of
special form radioactive material approved under 1973 Edition of Safety Series No. 6. In
SSR-6, 2018 Edition, special form radioactive material that was shown to meet the
provisions of the 1985 through 2012 Editions of IAEA standards may continue to be
used, with some additional restrictions on approval and fabrication. The IAEA’s SSR-6,
2018 Edition, prohibits fabrication of special form radioactive material that received
unilateral approval under the 1985 Edition of Safety Series No. 6 or 1985 (as Amended
1990) Edition of Safety Series No. 6. Also, after December 31, 2025, IAEA standards
prohibit new fabrication of special form radioactive material sources to a design that had
received unilateral approval under the 1996 Edition; 1996 Edition (Revised); 1996 (as
Amended 2003) Edition of TS-R-1; TS-R-1, 2005 Edition; TS-R-1, 2009 Edition; and
SSR-6, 2012 Edition.
Finally, in paragraphs 832–833, SSR-6, 2018 Edition, the IAEA revised the
package identification number in the CoC to delete the year designation (i.e., “-85” or
“-96”) for those package designs that are approved to SSR-6, 2018 Edition.
In the 2004 final rule (69 FR 3698; January 26, 2004), the NRC adopted the
following grandfathering provisions in § 71.19 for previously-approved packages:
ï‚·

Packages approved under NRC regulations that were compatible with the provisions
of the 1967 Edition of Safety Series No. 6 may be used for a 4-year period after
adoption of the final rule, presuming fabrication was completed by August 31, 1986;

ï‚·

Packages approved under NRC regulations that became effective on September 6,
1983 (see 48 FR 35600; August 5, 1983), which are compatible with the provisions
of the 1973 or 1973 (as amended) Editions of Safety Series No. 6, may no longer be

fabricated, but may still be used;
ï‚·

Packages approved under NRC regulations that are compatible with the provisions
of the 1985 or 1985 (as amended 1990) Editions of Safety Series No. 6, and
designated as "-85" in the package identification number, may not be fabricated after
December 31, 2006, but may still be used; and

ï‚·

Package designs approved under any pre-1996 IAEA standards (i.e., NRC packages
with an "-85" or earlier package identification number) may be resubmitted to the
NRC for review against the current NRC regulations. If the package design
described in the resubmitted application meets the current NRC regulations, the
NRC may issue a new CoC for that package design with a "-96" designation in the
package identification number.
In that same 2004 rulemaking, the NRC did not revise its grandfathering

provisions on special form radioactive material in § 71.4 because NRC regulations were
already consistent with the 1996 Edition of TS-R-1.
The NRC rulemaking in 2015 (80 FR 33988; June 12, 2015) made two minor
changes to the transitional arrangements regulations. First, the grandfathering provision
that was in § 71.19(a) for packages approved under NRC standards that were
compatible with the provisions of the 1967 Edition of Safety Series No. 6 was deleted
since that provision expired on October 1, 2008. Second, the definition of “special form
radioactive material” was revised to allow special form radioactive material that was
successfully tested using the current requirements of § 71.75(d) to continue to qualify as
special form radioactive material, if the testing was completed before September 10,
2015.
Consistent with past practices, the NRC is proposing transitional arrangements to
phase out older packages without a “-85” or “-96” in the package identification number,
and limit use of packages with a “-96” to those whose fabrication has been completed by
December 31, 2028, and consistent with DOT, limit fabrication of special form sources.
The NRC determined that it is appropriate to begin a phased discontinuance of these

older packages to further harmonize NRC’s regulations with the IAEA standards in SSR6, 2018 Edition. The DOT supports this discontinuation and coordinated with the IAEA
on the update to its standards. While the NRC has not identified safety issues that
necessitate the discontinuation of these older packages, they are no longer acceptable
in jurisdictions that use the IAEA requirements. The NRC views that the advantages of
consistent approvals across jurisdictions outweigh the value of retaining the
authorization for these packages. The approach being taken is consistent with the
NRC’s 2004 rulemaking. Given this experience, the NRC does not expect that certificate
holders will have challenges showing compliance with the regulations in effect at the
time the application is submitted for revision.
The NRC is proposing to revise its transitional arrangements to be consistent
with the IAEA, as follows:
1.

Phase out the use of packages approved to NRC regulations that were

harmonized with the IAEA’s 1973 Edition and 1973 (as Amended) Edition of Safety
Series No. 6, 8 years after the effective date of this rulemaking. These packages would
be required to be recertified, removed from service, or used via exemption.
2.

Prohibit the use of packages with a “-96” in the package identification

number for which fabrication of the packaging was completed after December 31, 2028,
and require multilateral approval (as defined in 49 CFR 173.403, “Definitions”) for
packages to be used for international shipment after December 31, 2025. Revise
§ 71.17(e) to state that packages with a “-96” in the package identification number would
become previously approved packages and subject to the current § 71.19(c).
3.

Coordinate with the DOT and make appropriate changes to § 71.4 to

align with the definition of “special form radioactive material” that the DOT is proposing
to adopt as part of their harmonization rulemaking, since DOT is the lead for certifying
special form sources. The NRC is proposing to allow continued use of special form
radioactive material that was approved to the regulations in effect from October 1, 2004
to the effective date of this rulemaking, provided they are fabricated on or before

December 31, 2025.
4.

Allow for package designs with a "-96" or earlier package identification

number to be resubmitted to the NRC for review against the current standards. If the
package design described in the resubmitted application meets the current standards,
the NRC may issue a new CoC for that package design without a year designation.
The NRC notes that the IAEA eliminated the approval year in the package
identification number for packages approved to SSR-6, 2018 Edition. Packages that
were approved to NRC regulations harmonized with the 1973 Edition of Safety Series
No. 6 do not have a year designation in the package identification number. To avoid
confusion regarding these older packages, the NRC would revise all existing CoCs that
do not have a “-85” or “-96” in their package identification number to add a provision that
those CoCs cannot be renewed beyond the end date of the 8-year phase out period
without being recertified to the revised version of 10 CFR part 71.
Issue 11. Inclusion of Head Space for Liquid Expansion
The NRC’s regulation in § 71.87, “Routine determinations,” requires that before
each shipment of licensed material, the licensee must ensure that the package, which
includes its contents, satisfies the applicable requirements of part 71. One such
requirement is that the licensee must determine in accordance with § 71.87(d) that any
system for containing liquid is adequately sealed and has adequate space or other
specified provision for expansion of the liquid.
The NRC’s requirement in § 71.87(d) is compatible with the DOT’s regulations at
49 CFR 173.24(h)(1), “General requirements for packagings and packages.” That
regulation requires: “When filling packagings and receptacles for liquids, sufficient
ullage (outage) must be left to ensure that neither leakage nor permanent distortion of
the packaging or receptacle will occur as a result of an expansion of the liquid caused by
temperatures likely to be encountered during transportation.”
The DOT’s regulations in 49 CFR 173.412(k), “Additional design requirements for
Type A packages,” contain a general design requirement for Type A packages designed

to contain liquids to ensure that packages provide for ullage to accommodate variations
in temperature of the contents. The term “ullage” refers to the unfilled space in a
container, or the amount by which the contents of a container fall short of being full.
Because DOT’s regulations for Type AF, Type B, and Type BF packages refer to the
NRC’s regulations, DOT’s regulations do not contain design requirements for Type AF,
Type B, or Type BF packages. Type A, Type AF, Type B, and Type BF packages are
defined in § 71.4, “Packages.”
The IAEA standards in paragraph 649, SSR-6, 2018 Edition, require that “The
design of a package intended for liquid radioactive material shall make provision for
ullage to accommodate variations in the temperature of the contents, dynamic effects
and filling dynamics.”
The NRC regulations have an operational requirement in § 71.87(d) to ensure
that for a system containing liquid, there is sufficient head space, or other specified
provision to accommodate the expansion of liquid. The NRC does not, however, have a
comparable design requirement for Type AF and Type B packages in 10 CFR part 71 to
that in DOT’s regulations. Even though the NRC’s regulations do not include a
comparable design requirement for ensuring sufficient space to allow for liquid
expansion, any Type AF or Type B package design certified by the NRC must comply
with § 71.87 and DOT regulations in 49 CFR 173.24(h) on ullage when being filled.
During review of applications for either a new CoC or an amendment to an
existing CoC, the NRC reviews whether the requirements in § 71.87(d) are reflected in
the operating procedures for packages with liquid contents. Each package approval
issued by the NRC contains a condition to ensure that the package is prepared in
accordance with the operating procedures in the application. This ensures that all
package users, whether NRC licensees or not, comply with the requirements listed in
§ 71.87, as appropriate for the package design.
Although the NRC regulations ensure that adequate ullage exists, the NRC has
received on occasion an application that did not evaluate whether there was sufficient

design space in a container with liquids. To clarify this requirement, the NRC is
proposing to revise § 71.43, “General standards for all packages,” to add a design
requirement for a package designed to contain liquids to ensure adequate ullage during
evaluation of the tests and conditions for normal conditions of transport and hypothetical
accident conditions.
Issue 12. Revision of Quality Assurance Program Biennial Reporting Requirements
On June 12, 2015, the NRC issued a final rule (80 FR 33988), updating the
administrative procedures for the QAP requirements described in 10 CFR part 71,
subpart H, “Quality Assurance.” Specifically, the NRC added § 71.106 to establish
requirements for QAP changes and associated reporting requirements.
Previously, all changes made to QAP approvals had to be reviewed and
approved by the NRC before they could be implemented. The provisions in § 71.106
allow changes to QAPs that do not reduce commitments, such as those that involve
administrative improvements and clarifications, spelling corrections, and non-substantive
changes, to be made and implemented without prior NRC approval. QAP changes that
would reduce commitments require prior NRC approval.
In addition, § 71.106 requires that changes to QAPs that do not reduce
commitments must be submitted to the NRC every 24 months. That final rule also
specified, “If a quality assurance program approval holder has not made any changes to
its approved quality assurance program description during the preceding 24-month
period, the approval holder will be required to report this to the NRC” (80 FR 33994). In
addition, the NRC’s guidance document for 10 CFR part 71 QAPs, Regulatory Guide
7.10, Revision 3, was updated in conjunction with the 2015 final rule to state that if no
changes were made to the QAP, a QAP approval holder would indicate to the NRC that
no changes were made.
The requirement for a report, even if no changes were made during the
preceding 24-month period, is necessary as the NRC inspection program for 10 CFR
part 71 QAP approval holders relies on having current information about the QAP

available to the NRC. The NRC considers the 24-month reporting requirement, including
when no changes are made, as providing an appropriate balance between the burden
placed on the QAP approval holders and the need to ensure that the NRC has current
information for its oversight of these QAPs. Most QAP approval holders subject to
periodic inspection are inspected every 5 years or on an as-needed basis. Another
benefit to receiving a report even when no QAP changes have been made is that the
QAP reporting requirements in 10 CFR part 71 would be consistent with those in
§§ 50.54(a)(3) and 50.71(e)(2) for 10 CFR part 50 QAPs. Since the 2015 final rule
became effective, the NRC has received questions and concerns from industry on this
subject since the language in § 71.106 does not state that QAP approval holders must
report even if there were no changes in the prior 24-month period.
The NRC is proposing to revise § 71.106(b) to clarify that a biennial report must
be submitted to the NRC even if no changes are made to the QAP during the reporting
period.
Issue 13. Deletion of Type A Package Limitations in Fissile Material General Licenses
The general license criteria in § 71.22 allow NRC licensees to ship small
quantities of fissile material in packages that have been assigned a criticality safety
index (CSI) to ensure accumulation control for packages on a conveyance. The
provisions of § 71.22 require that 1) the fissile material is in a Type A package that
meets the requirements of 49 CFR 173.417(a); 2) licensees have an NRC-approved
QAP satisfying the provisions of 10 CFR part 71, subpart H; 3) there is no more than a
Type A quantity of radioactive material; 4) there is less than 500 grams total of beryllium,
graphite, or hydrogenous material enriched in deuterium; and 5) the package is labeled
with a CSI that meets the limits in § 71.22(d). The regulation in § 71.22(e)(1) provides
an equation to calculate package CSI:
grams of 233U
grams of Pu
grams of 235U
CSI = 10
+
+
Z
X
Y
where X, Y, and Z are mass limits of 235U, 233U, and plutonium obtained from Table 71-1
(if 233U or plutonium are present) or Table 71-2.

Similarly, the general license criteria in § 71.23 allow NRC licensees to ship small
quantities of special form plutonium in packages that have been assigned a CSI to
ensure accumulation control for packages on a conveyance. The provisions of § 71.23
require that 1) the fissile material is in a Type A package meeting the requirements of 49
CFR 173.417(a); 2) licensees have an NRC-approved quality assurance program
satisfying the provisions of 10 CFR part 71, subpart H; 3) there is no more than a Type A
quantity of radioactive material; 4) there is less than 1,000 grams of plutonium, provided
that the total amount of 239Pu and 241Pu constitutes less than 240 grams of the plutonium
in the package; and 5) the package is labeled with a CSI that meets the limits in
§ 71.23(d). The regulation in § 71.23(e)(1) provides an equation to calculate package
CSI:

𝐶𝑆𝐼 = 10

𝑔𝑟𝑎𝑚𝑠 𝑜𝑓 239𝑃𝑢 + 𝑔𝑟𝑎𝑚𝑠 𝑜𝑓 241𝑃𝑢
The calculations that support the mass limits in § 71.22 include conservative
assumptions regarding neutron moderation and water reflection, i.e., optimally
moderated spheres of 235U, 233U, and 239Pu with full water reflection. The mass limits in
§ 71.23 have a similar basis, but are higher for the two fissile plutonium isotopes, as the
material is special form and will not redistribute significantly. In both cases, it is
assumed that the material will remain in the package under normal conditions of
transport because of the Type A package requirement but can reconfigure outside of the
package under hypothetical accident conditions. The limitation to a Type A quantity of
radioactive material in a Type A package, however, is not consistent with the mass limits
for some fissile nuclides in some cases (e.g., the mass limits for 239Pu in Table 71-1 are
37 grams or 24 grams, depending on the degree of moderation, while the A2 value for
239Pu

is equivalent to 0.435 grams). In addition, the requirement in § 71.23 does not

consistently refer to “special form sealed sources” in that paragraph (a) also refers to PuBe sealed sources. While all special form sources are sealed sources, not all sealed

sources meet the definition of special form material in 10 CFR 71.4.
Removing the limitation to a Type A quantity of radioactive material in a Type A
package would allow licensees to ship material under the general licenses in §§ 71.22
and 71.23 in a Type B package. When shipping material that meets the mass limits of
the general licenses in §§ 71.22 and 71.23 in a Type B package, the criticality safety
conclusions associated with these mass limits remain valid. In fact, the material would
be less likely to present a criticality hazard, as Type B packages generally are more
robust and have more mass, which would increase neutron absorption, limit releases
under hypothetical accident conditions, and prevent material from multiple packages
from redistributing together under optimum moderation conditions.
Revising the general licenses to authorize transport in a Type B package would
also require conforming changes to § 71.0(d)(1). The regulations in § 71.0(d)(1) state
that use of the general licenses in § 71.22 or § 71.23 does not require NRC approval.
Package approval is not currently required by the NRC because the conditions of the
general licenses require the contents to be in a Type A package. The regulations in
§ 71.14(b)(1) exempt the licensee from all requirements in 10 CFR part 71, except for
§§ 71.5 and 71.88, when shipping a Type A quantity. Because the NRC is proposing to
revise §§ 71.22 and 71.23 to authorize shipment of a Type B quantity of radioactive
material, an NRC package approval would be required for shipment of the Type B
quantity of radioactive material. The NRC package approval for the Type B quantity of
radioactive material would not include evaluation of criticality safety because the
criticality safety is assured for shipment of fissile material authorized under one of these
general licenses.
While NRC is not proposing to revise §§ 71.22(b) and 71.23(b), which require
that the licensee have an NRC-approved QAP. Applications for QAP approvals use a
graded approach, based on the planned activities and shipments that a licensee plans to
make. For example, if a licensee has a QAP that was approved for making only Type A
shipments under § 71.22 or § 71.23, then the licensee would need to obtain additional

NRC approval for a QAP that includes QA items necessary for making Type B
shipments.
In addition, because the NRC is proposing to authorize shipments of Type B
packages in §§ 71.22 and 71.23, the NRC is proposing to include three new paragraphs
in §§ 71.22 and 71.23 that are similar to the requirements in § 71.17(c), (d), and (e).
The NRC is proposing to add a new requirement in §§ 71.22(f) and 71.23(f) to ensure
that, for shipments made using the respective general license, each licensee must
comply with § 71.17(c), i.e., the licensee must: 1) maintain a copy of the NRC approval,
including all referenced documents; 2) comply with the terms and conditions of the NRC
approval and the applicable requirements of subparts A, G, and H in 10 CFR Part 71;
and 3) prior to first use, register to use the package. A licensee is only required to
register once to use a package, and therefore a licensee already registered to use the
package via § 71.17 would not have to re-register to use the package under one of these
two general licenses.
The NRC is proposing to add a new requirement in §§ 71.22(g) and 71.23(g) to
state that, for a package to be used under the respective general license, the NRC
package approval must state that the package can be used under the general license in
either § 71.17 or the general license in § 71.22 or § 71.23. Authorizing use under the
general license in § 71.17 would ensure that existing, approved Type B package designs
could also be used to transport the material authorized by one of the two general
licenses in § 71.22 or § 71.23.
Finally, the NRC is proposing to add a new requirement in §§ 71.22(h) and
71.23(h) to ensure that any Type B package used under the respective general license
approved by the NRC before the effective date of the final rule is subject to the
transitional arrangements in § 71.19. Issue 10 in Section III of this document describes
the NRC’s proposed changes to its transitional arrangements.
In summary, the NRC is proposing to remove the restriction in §§ 71.22 and
71.23 to ship Type A material in only a Type A package (i.e., allowing shipment of

material up to the mass limits in a Type B package); to add three new paragraphs in
§§ 71.22 and 71.23; and to make conforming changes to § 71.0(d)(1). Additionally, the
NRC is proposing to clarify that only special form sealed sources, not just sealed
sources may be delivered to a carrier for transport using the general license in § 71.23.
Issue 14. Deletion of 233U Restriction in Fissile General License
The general license criteria in § 71.22 allow NRC licensees to ship small
quantities of fissile material in packages that have been assigned a CSI to ensure
accumulation control for packages on a conveyance. General license users assign a
CSI based on the equation in § 71.22(e)(1), and the fissile mass limits in either Table 711 or 71-2 to 10 CFR part 71. Table 71-2 contains mass limits for shipping uranium
enriched to various weight percent levels in 235U. However, § 71.22(e)(5) states in part
that the lower mass values of Table 71-1 must be used if the enrichment level of
uranium is unknown, if the amount of plutonium exceeds one percent of the mass of
235U,

or if 233U is present in the package.
While 233U is not present in natural uranium, it may be present in very low

concentrations in some facilities that may have handled 233U in the past. These
contamination-level concentrations, while detectable with modern isotopic assay
methods and physically “present,” are not important for criticality safety of 235U
transportation. The calculations used to support the enrichment limit for § 71.15(d), for
up to 1.0 weight percent enriched uranium, demonstrate that this limit is safe provided
the plutonium and 233U are limited to less than one percent of the mass of 235U. The
same limitation could be applied to the use of Table 71-2 limits for shipping enriched
uranium under § 71.22, without affecting criticality safety.
The NRC is therefore proposing to revise § 71.22 to limit the 233U to less than
one percent of the mass of 235U, similar to the provision limiting plutonium in
§ 71.22(e)(5)(ii).
Issue 15. Other Recommended Changes to 10 CFR Part 71

As described in the draft regulatory basis, Issue 15 groups several topics
identified by the NRC, some of which are not directly related to harmonizing NRC
requirements with IAEA standards, and include clarifications to ensure compatibility with
the DOT and clarifications to Agreement State regulations.
Issue 15.1. Deletion of Duplicative Reporting Requirements
In the 2002 proposed rule (67 FR 21390, April 30, 2002), the NRC proposed
changes to its reporting requirements in § 71.95, “Reports.” Those proposed changes
would have: 1) required licensees to obtain certificate holder input before submitting an
event report; 2) provided direction on the content of the written report; and 3) lengthened
the reporting requirement date to 60 days, consistent with other reporting requirements
in NRC regulations. The proposed rule recommended adding 71.95(a)(1) and (2) and
71.95(b), but not the current 71.95(a)(3).
In the final rule (69 FR 3697, January 26, 2004), the NRC stated that the
proposed rule had inadvertently left out new paragraph (a)(3), mentioned in the
proposed rule’s regulatory analysis, that would retain the existing requirement for
licensees to report instances of failure to follow the conditions of the CoC while a
packaging was in use. Paragraph (a)(3) was thus added to the final rule. However, in
adding that paragraph to the final rule, the NRC introduced duplicative language
between it and paragraph (b).
The NRC is proposing to delete the duplicative text in paragraph (a)(3).
Issue 15.2. Revision of the Definition of Low Specific Activity
The NRC is proposing to modify the first sentence in the definition of “Low
Specific Activity (LSA) material” in § 71.4 to change “excepted under § 71.15” to
“exempted under § 71.15.” This change would make the definition of LSA in § 71.4
consistent with the title of § 71.15, “Exemption from classification as fissile material” and
ensure that it is clear that LSA packages may contain fissile material up to the exemption
limits in § 71.15.
Issue 15.3. Revision of Tables Containing A1 and A2 Values and Exempt Material

Activity and Consignment Limits
The IAEA has made changes in SSR-6, 2018 Edition, related to the A1 and A2
activity values and the exempt material activity concentrations and exempt consignment
activity limits. The DOT is the lead agency for information related to the A1 and A2
values and for the exempt material activity concentrations and exempt consignment
activity limits, as provided in 49 CFR 173.435 and 173.436, respectively. The NRC has
corresponding information in 10 CFR part 71, Appendix A, Tables A-1 and A-2.
To be considered radioactive material under DOT’s regulations (i.e., Class 7
(radioactive) material as defined in 49 CFR 173.403), the material must exceed both the
nuclide specific exemption concentration limit and the consignment exemption activity
limit. The A1 and A2 values are quantities of radioactivity that are used in the
transportation regulations to determine the type of packaging necessary for a particular
radioactive material shipment. Each radionuclide is assigned an A1 and an A2 value,
where A1 is the maximum activity of special form material that is permitted in a Type A
package, and A2 is the maximum activity of normal form radioactive material that is
permitted in a Type A package as prescribed in 10 CFR 71.4 and 49 CFR 173.403. The
NRC’s and the DOT’s transportation regulations include package activity limits based on
fractions or multiples of the A1 and A2 values (e.g., 10-3A2 and 3,000A2, respectively).
In its concurrent harmonization rulemaking, the DOT is proposing to make
changes to 49 CFR 173.435, “Table of A1 and A2 values for radionuclides,” and 173.436,
“Exempt material activity concentrations and exempt consignment activity limits for
radionuclides,” by adding seven radionuclides, including barium-135m, germanium-69,
iridium-193m, nickel-57, strontium-83, terbium-149, and terbium-161. The NRC is
proposing to make corresponding changes to Tables A-1 and A-2 to add these
radionuclides. The NRC is proposing to revise the specific activity of natural rubidium
(Rb(nat)) to correct an error that was introduced in the 1995 version of the rule. Table
A-1 of Appendix A to 10 CFR part 71 gives the specific activity as 6.7×106 TBq/g,
1.8×108 Ci/g. However, the correct value for the specific activity of Rb(nat) is 670 Bq/g

(6.7×10-10 TBq/g, 1.8×10-8 Ci/g). The A1 and A2 values were not impacted by this error
and remain correct. The NRC is also proposing to revise footnote c at the end of Table
A-2 to state that in the case of thorium-natural, the parent radionuclide is thorium-232,
and in the case of uranium-natural, the parent radionuclide is uranium-238. Further, the
NRC is proposing to editorially revise several other radionuclides to move the name of
the element and its atomic number (shown in the second column of each table) to the
first instance of that element alphabetically in the tables.
Issue 15.4. Revision to Agreement State Compatibility Categories
The NRC is proposing several changes to the compatibility category designations
related to the QAP and reporting requirements. These changes would ensure that
Agreement States have the appropriate authority to approve, inspect, and enforce QAPs
for their licensees, as well as that the NRC and Agreement States receive important
reports regarding issues with radioactive material shipments.
The NRC is proposing to revise the compatibility category designations for the
regulations containing QAP requirements for those Agreement States that have
licensees located within their States who use NRC-approved Type B packages, other
than for industrial radiography, to ship Type B quantities of radioactive material; or have
licensees that ship using the general license in § 71.21, “General license: Use of foreign
approved package”; § 71.22, “General license: Fissile material”; or § 71.23, “General
license: Plutonium-beryllium special form material.” The NRC is also proposing to revise
the compatibility category designation for the reporting requirements in § 71.95.
In the 2004 final rule (69 FR 3697; January 26, 2004) that revised § 71.101,
“Quality assurance requirements,” the NRC stated that § 71.101(b), and (c)(1) are
designated as Compatibility Category C for those Agreement States that have licensees
that use Type B packages, other than for industrial radiography. For Compatibility
Category C, the essential objectives of the NRC program elements should be adopted
by such Agreement States. The NRC is proposing to change the compatibility category
designation for 71.101(b) and (c)(1) from C to B. This is consistent with Management

Directive 5.9, “Adequacy and Compatibility of Program Elements for Agreement State
Programs,” which states that program elements in Compatibility Category B are those
that apply to activities that cross jurisdictional boundaries. Since the QAP activities in
71.101(b) and (c)(1) are used during domestic shipping of radioactive material and
therefore cross jurisdictional boundaries, a B compatibility would align with Management
Directive 5.9 criteria. Also, many of the regulations that contain QAP review criteria
(e.g., §§ 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, and 71.125)
were addressed in the 2004 rule, but were designated as Compatibility Category NRC,
which relate to areas of regulation reserved to the NRC that cannot be adopted by the
Agreement States. The NRC is proposing to address these compatibility issues in this
proposed rule so that, consistent with the intent of the 2004 rulemaking, Agreement
States can adopt compatible QAP regulations that would require their licensees to follow
these QAP criteria and allow Agreement States to approve, inspect and enforce their
licensees’ QAPs. Specifically, this rule proposes to correct the compatibility category
designation to B for many of these regulations that are currently Compatibility Category
NRC, C, or D. This change would require Agreement States to have essentially identical
regulations and would give the Agreement States the authority to approve, inspect and
enforce their licensees’ QAPs. Only Agreement States with licensees that use Type B
packages, other than for industrial radiography, or with licensees that ship using the
general license in § 71.21, § 71.22, or § 71.23, which also requires an approved QAP,
would be impacted.
Additionally, the regulations in § 71.95 require NRC licensees to submit a written
report to the NRC of instances in which there is a significant reduction in the
effectiveness of any NRC-approved package; details of defects with safety significance
in any NRC-approved package, after first use; and instances in which the conditions of a
CoC were not followed during shipment. In the 2004 final rule (69 FR 3697; January 26,
2004) that revised § 71.95, the NRC stated that the compatibility category for § 71.95 is
Category D; therefore, it does not need to be adopted by the Agreement States to be

compatible with the NRC’s regulatory program. The reporting requirements in § 71.95(a)
are to ensure that the NRC is alerted to instances in which a package may have a defect
or has a significant reduction in effectiveness such that, as needed, other licensees
authorized to use the package are made aware of the possible issues. Agreement State
licensees also use NRC-approved packages, including industrial radiography devices,
but are not subject to any of the requirements in § 71.95 and, therefore, are not required
to submit a report to the NRC pursuant to § 71.95. The NRC is proposing to change the
compatibility category for § 71.95(a) to Compatibility Category C in order to have
Agreement State regulations require notification to the NRC of these instances. This will
clarify that if a State licensee uses an NRC-approved package that has a defect or has a
significant reduction in effectiveness the NRC is aware such that others using the
package can be made aware of the situation. The NRC also is proposing to update the
compatibility category for § 71.95(b) to Compatibility Category C to ensure that the
Agreement State agency receives these reports from its licensees indicating instances
when the CoC was not followed. As noted in the 1995 final rule (60 FR 50248, 50259),
the purpose of this requirement is to provide feedback on QAP effectiveness.
Consistent with the compatibility category corrections for other QAP related regulations,
this proposed rule would also correct the compatibility category for § 71.95(b) so that
Agreement States receive these QAP-related reports. The compatibility categories for
§ 71.95 (c) and (d) would also be revised to Compatibility Category C so that these
reports contain the required information.
In summary, the NRC is proposing to revise the compatibility category for 1)
§ 71.101(b) and (c)(1) from a Compatibility Category C to B to be in alignment with the
criteria in Management Directive 5.9; 2) many of the QAP-related regulations (e.g.,
§§ 71.109, 71.111, 71.113, 71.115, 71.117, 71.119, 71.121, 71.123, and 71.125) from a
Compatibility Category NRC, C, or D to a B to allow the Agreement States the authority
to approve, inspect and enforce these regulations; and 3) the reporting requirements in
§ 71.95(a) and (b) from a Compatibility Category D to C so that the NRC receives

reports from Agreement State licensees on package defects pursuant to § 71.95(a), and
that Agreement State regulators receive reports when their licensees do not use an
NRC-approved package in accordance with the CoC pursuant to § 71.95(b), and to
§ 71.95(c) and (d) so that these reports contain the required information.
Issue 15.5. Deletion of Redundant Advance Notification Requirements for
Shipment of Spent Nuclear Fuel
Section 71.97 is titled “Advance notification of shipment of irradiated reactor fuel
and nuclear waste.” However, advance notification requirements for irradiated reactor
fuel (and, equivalently, spent nuclear fuel) are separately included in the more general
requirements of 10 CFR part 73, “Physical protection of plants and materials.”
Specifically, as required in § 73.37(b)(2), licensees are required to provide advance
notification of shipment to the Governor of a State and/or Tribal official for any shipment
crossing the State or Tribal boundary when the shipment contains greater than 100
grams irradiated reactor fuel and the external radiation dose rate is greater than 1 Gy
(100 rad) per hour at a distance of 1 meter (3.3 feet) from any accessible surface without
intervening shielding. Licensees are also required to provide notification of such
shipments to the NRC in accordance with § 73.72. Additionally, as required in § 73.35,
“Requirements for physical protection of irradiated reactor fuel (100 grams or less) in
transit,” licensees who transport 100 grams or less of irradiated reactor fuel, when the
external radiation dose rate is greater than 1 Gy (100 rad) per hour at a distance of 1
meter (3.3 feet) from any accessible surface without intervening shielding, are required
to provide advance notification of shipment in accordance with § 37.77. When 10 CFR
part 37 was established in 2013, this requirement was introduced, but the “irradiated
reactor fuel” aspect was not removed from § 71.97. Therefore, licensees may need to
produce two reports for a single shipment to meet the advance notification requirements
of §§ 71.97 and 73.37 or § 73.35. To address this potential inefficiency the NRC is
proposing to modify § 71.97 to remove references to irradiated reactor fuel.

IV.

Specific Request for Comment

The NRC is seeking comment and feedback from the public on this proposed
rule. The NRC is particularly interested in comment and supporting rationale from the
public on the following:
QUESTION 1: IAEA changes in SSR-6 (2018 Edition) not in the scope of this
proposed rule
Starting in 2016, while developing the regulatory basis for this proposed rule, the
NRC considered the changes in SSR-6, 2012 Edition, and the proposed changes that
were being considered for SSR-6, 2018 Edition, which were eventually issued in June
2018. The NRC contracted with Oak Ridge National Laboratory (ORNL) to develop
ORNL/TM-2014/658, “Comparison of the International and United States Domestic
Radioactive Material Transport Regulations.” In this document, ORNL compared both
NRC and DOT regulations to SSR-6, 2012 Edition, and noted the differences. The NRC
then compared the changes between SSR-6, 2018 Edition, and the 2012 Edition to
determine which changes affect NRC regulations and whether those changes should be
included in this proposed rule. Based on this review, the NRC did not include the
following IAEA changes in the scope of this proposed rule:
1.

Issue 1 consisted of four different sub-issues: Issue No. 1a: New Fissile

Exceptions in IAEA SSR-6, paragraph 417; Issue No. 1b: Competent AuthorityApproved Fissile Exception, SSR-6, paragraph 417(f); Issue No. 1c: CSI-Controlled
Fissile Material Packages, SSR-6, paragraph 674; and Issue No. 1d: Plutonium
Shipments in Type A Packages, SSR-6, paragraph 675.
For issue 1a, the NRC considered whether to adopt the fissile exceptions in
paragraphs 417(c), without consignment limits in paragraph 570(c); the consignment
limit in paragraph 570(d) associated with the package mass limit in paragraph 417(d);
and the exception in paragraph 417(e) and its associated exclusive use restriction in
paragraph 570(e), but with a mass limit of 140 g instead of the IAEA mass limit of 45

grams of fissile material from SSR-6, 2018 Edition, into the NRC regulations. The NRC
chose not to adopt the consignment limits in 570(c) and (d) for the fissile exceptions in
417(c) and 417(d), respectively because consignment limits do not prevent the
accumulation of packages on a transport conveyance, as there is no limit to the number
of consignments that may be present on a single conveyance. Additionally, the
accumulation on a single conveyance of the number of these packages required to
approach criticality is not credible.
After evaluation of Issue 1b, the NRC is not proposing to add the new “competent
authority-approved” fissile exception in paragraph 417(f) into the NRC regulations. If an
NRC licensee wished to ship a material that did not meet the fissile material exemption
or general license criteria in 10 CFR part 71, and for which demonstration of
subcriticality in a package per the requirements of §§ 71.55 and 71.59 is deemed too
burdensome, the licensee could request a specific exemption under § 71.12. The NRC
notes that if an NRC licensee submitted a “competent authority-approved” exception, the
approval would include both NRC and DOT reviews and issuance of the exception and
the NRC review and findings would be similar to those of either an exemption or NRCissued CoC.
After evaluation of Issue 1c, the NRC is not proposing to add CSI-controlled
fissile material packages that the IAEA incorporated into SSR-6, paragraph 674. The
IAEA SSR-6, paragraph 674(a), contains fissile material mass limits (per Table 13 in
SSR-6, paragraph 674) and a CSI determination for packages with a minimum external
dimension of 10 centimeters, which are not required to withstand normal conditions of
transport in SSR-6, paragraphs 719–724. The IAEA SSR-6, paragraph 674(b), contains
similar fissile material mass limits, and a formula for determination of a lower CSI, for
packages which withstand normal conditions of transport while maintaining a larger
minimum external dimension of 30 centimeters. The IAEA SSR-6, paragraph 674(c),
contains the same CSI calculation as paragraph 674(b), for packages that withstand

normal conditions of transport while maintaining a minimum external dimension of
10 centimeters, with a limit of 15 grams fissile material per package.
The NRC does not propose to adopt the changes in IAEA SSR-6, paragraph 674,
because the NRC has determined that the mass limits and other requirements in
§§ 71.22 and 71.23 are appropriate for providing criticality safety equivalent to packages
approved under the criticality safety requirements of §§ 71.55 and 71.59. Adopting the
provisions of IAEA SSR-6 would result in more restrictive mass limits for the fissile
material general licenses authorized under 10 CFR part 71.
The NRC evaluated issue 1d, SSR-6, paragraph 675, to add NRC requirements
for shipment of plutonium in a nonfissile package, with accumulation control provided by
the calculation of a CSI. This provision was included in SSR-6, 2012 Edition but without
accumulation control. The NRC’s fissile exemption in § 71.15(f) is similar in that it limits
the package to 1000 g of plutonium, of which not more than 20 percent by mass may be
plutonium-239, plutonium-241, or any combination of the two; however, the NRC
regulation does not include accumulation control via a CSI calculation. The NRC has
determined that the fissile exemption in § 71.15(f) is safe without accumulation control,
and that there is no safety benefit to limiting accumulation through the use of a CSI, in
order to be consistent with the IAEA standards. Therefore, the NRC is not proposing to
harmonize with paragraph 675, SSR-6, 2018 Edition.
2.

The NRC considered adopting the reduced external pressure value of

60 kPa from paragraph 645 and the air transport package requirements from paragraph
621. The NRC is not proposing to harmonize with paragraphs 621 and 645, SSR-6,
2018 Edition, as discussed for Issue 2 in Section III of this proposed rule, to avoid
creating unnecessary mode-specific restrictions within 10 CFR part 71.
3.

Inclusion of Type C Package Standards (paragraphs 669–672) – The

NRC considered adding Type C package standards for domestic transport, but there
was not an expressed need for domestic transport of packages approved to Type C

standards. Therefore, the NRC is not proposing to add Type C package standards in
this proposed rule.
4.

Testing and reporting the integrity of the containment system and

shielding, and assessing criticality safety (paragraph 716), and additional description of
the impact of the tests on packages (paragraphs 718–737) – The NRC reviewed its
regulations for an application for approval of a package design and considered its
regulations sufficient to obtain the information needed to determine whether a package
design meets the requirements in 10 CFR part 71.
5.

Addition of LSA Fissile Shipments (paragraphs 518, 519, 520) – Since

LSA packages are self-certified under DOT regulations, other than the fissile material
exemptions (§ 71.15) and fissile material general licenses (§§ 71.22 and 71.23), there is
no mechanism for adding fissile material to an LSA package without NRC approval.
Under current NRC regulations, the package could be certified but would become a
Type BF or Type AF package, depending on the quantity of radioactive material in the
package, and therefore the NRC did not consider any revision necessary.
6.

Safety Factors for Lifting Attachments (paragraph 608) – The NRC

regulations in § 71.45 contain quantitative criteria for evaluating lifting attachments that
are considered a structural part of the package. The IAEA standards state an
“appropriate” safety factor must be used. In its review, the NRC determined that
adopting the IAEA changes would not result in safety benefits beyond those in § 71.45.
7.

Shipment after Storage and Gap Analysis (paragraphs 503(e) and 809(k))

– The IAEA added regulations both for shipment after storage and a gap analysis for
packages in storage prior to shipment. The regulations in SSR-6, paragraph 503(e),
require that during storage, packages are maintained to ensure that all relevant
transportation standards in SSR-6 and certificates of approval for those packages will be
fulfilled. The NRC is not proposing to adopt paragraph 503(e) because, during its review
of packages for which storage is expected prior to transport (i.e., dual purpose casks or
canisters), the NRC ensures that the evaluations, operating procedures, maintenance

program and acceptance tests for transport take storage into consideration. In addition,
for any package that is stored prior to transport, existing NRC requirements (§§ 71.17(c)
and 71.87(b)) ensure that, prior to transport, the licensee must comply with the terms
and conditions of the NRC approval for the package design and ensure the package is in
unimpaired physical condition. Following the operating procedure, maintenance
program, and acceptance tests in the application is a condition of approval in all NRCapproved CoCs.
The NRC is not proposing to adopt paragraph 809(k), which requires “periodic
evaluation of changes of regulations, changes in technical knowledge and changes of
the state of the package design during storage.” The NRC’s transitional arrangements
authorize continued use of package designs approved to prior versions of the NRC
regulations, with limitations on fabrication and restrictions on modifications to package
designs without the need to demonstrate full compliance with the revised regulations.
Package designs compliant with the existing regulations do not become "unsafe" when
the regulations are revised (unless a significant safety issue is corrected in the revision).
If a significant safety issue is corrected in a rulemaking, NRC certificate holders for that
package design or type of package would be informed via generic communication (e.g.,
regulatory information summary, bulletin, or generic letter), and as appropriate, required
to take action, prior to a potential rule change. In addition, as stated previously, prior to
transport the licensee must comply with the terms and conditions in the NRC approval
and ensure the package is in unimpaired physical condition.
ï‚·

Is there anything in SSR-6, 2018 Edition, that the NRC did not include in the

scope of this proposed rule, but should have? In your comment, please explain why the
NRC should consider adding the change to the final rule and the associated benefits.
QUESTION 2: Removing Tables A-1 through A-4 in Appendix A to 10 CFR Part
71
The NRC transportation regulations in 10 CFR part 71 include appendix A to 10
CFR part 71, “Determination of A1 and A2.” The introductory material in paragraphs I–V

to appendix A includes information related to determining A1 and A2 values. Appendix A
includes four tables:
-

Table A-1: “A1 and A2 Values for Radionuclides”

-

Table A-2: “Exempt Material Activity Concentrations and Exempt Consignment
Activity Limits for Radionuclides”

-

Table A-3: “General Values for A1 and A2”

-

Table A-4: “Activity-Mass Relationships for Uranium”
The Secretary of Transportation has the authority to regulate the transportation of

hazardous materials per the Hazardous Materials Transportation Act, as amended and
codified in 49 U.S.C. 5101, et seq. The Secretary is authorized to issue regulations to
implement the requirements of the statute. The DOT’s Pipeline and Hazardous
Materials Safety Administration has been delegated the responsibility for the hazardous
materials regulations, which are contained in 49 CFR parts 100–185. These regulations
include the requirements for Class 7 (radioactive) material.
The DOT maintains the same information in 49 CFR 173.433 through 49 CFR
173.436 as found in the NRC’s appendix A to 10 CFR part 71. With the authority to
regulate the transportation of hazardous materials, including Class 7 (radioactive)
material, DOT is the lead agency for determining the basic radionuclide values (A1 and
A2 values) and the exempt material activity concentrations and exempt consignment
activity limits for radionuclides that are used in radioactive material transportation
activities. The DOT regulations include:
-

49 CFR 173.433, “Requirements for determining basic radionuclide values, and
for the listing of radionuclides on shipping papers and labels”

-

49 CFR 173.433, Table 7, “General Values for A1 and A2”

-

49 CFR 173.433, Table 8, “General Exemption Values”

-

49 CFR 173.434, “Activity-mass relationships for uranium and natural thorium”

-

49 CFR 173.435, “Table of A1 and A2 values for radionuclides”

-

49 CFR 173.436, “Exempt material activity concentrations and exempt
consignment activity limits for radionuclides”
The NRC recognizes challenges associated with maintaining the accuracy and

consistency of all the information in appendix A to 10 CFR part 71 with the parallel
information in 49 CFR chapter I, considering, in part, the periodic updates the DOT
makes to these regulations to harmonize with IAEA standards. Therefore, to minimize
duplicative information within the domestic transportation regulations, and to recognize
the DOT’s authority to regulate Class 7 (radioactive) material, the NRC is considering
removing the content of appendix A to 10 CFR part 71. Where it is necessary within the
subparts of 10 CFR part 71, the NRC would remove all references in 10 CFR chapter I
to information in appendix A to 10 CFR part 71 and replace those with references to the
appropriate regulation in 49 CFR chapter I.
ï‚·

Please comment on whether the NRC should consider removing Tables

A-1 through A-4 in appendix A to 10 CFR part 71 and instead refer to the appropriate
DOT tables in 49 CFR chapter I, rather than updating Tables A-1 through A-4 in
appendix A to 10 CFR part 71 as currently shown in this proposed rule. If so, would
there be a benefit to members of the public, including applicants and licensees? Please
explain your rationale.
QUESTION 3: Merits of requiring a biennial report for no changes to a QAP
As described in Section III of this document, in Issue 12, the NRC is proposing to
revise § 71.106 to achieve NRC’s stated intent in the 2015 final rule. Specifically, the
NRC is proposing to revise § 71.106(b) to clarify that a biennial report must be submitted
to the NRC even if no changes are made to the QAP during the reporting period. This
proposed requirement would benefit the NRC’s regulatory oversight of QAP approval
holders. The NRC inspection program for 10 CFR part 71 QAP approval holders relies
on having current information about the QAP available to the NRC, including the
reporting of no changes. The 24-month reporting period aims to provide an appropriate
balance between the burden placed on the QAP approval holders and the need to

ensure that the NRC has current information, especially when considering most QAP
approval holders subject to periodic inspection are inspected every 5 years or on an asneeded basis. Another benefit is that the revised QAP reporting requirements in 10 CFR
part 71 would be consistent with those in 10 CFR 50.54(a)(3) and 50.71(e)(2) for 10
CFR part 50 QAPs. The benefits and costs of the proposed requirement are described
in the regulatory analysis and the NRC estimates that the cost of compliance is very
small. The NRC is interested in the public’s feedback as to the benefits and costs of
requiring a no-change biennial report.
ï‚·

Please comment on the benefits and costs of requiring a 10 CFR part 71

QAP approval holder to submit a biennial report to the NRC even if no changes are
made to the QAP during the reporting period.

V.

Section-by-Section Analysis

The following paragraphs describe the specific changes in this proposed rule.

Section 71.0, Purpose and scope.
This proposed rule would revise paragraph (d)(1) to clarify general license
package approval requirements.

Section 71.4, Definitions.
This proposed rule would revise the definitions for Low Specific Activity material,
Special form radioactive material, and Surface Contaminated Object, delete the
definition for Low Specific Activity-III Leaching Test, and add a new definition for
Radiation level.

Section 71.15, Exemption from classification as fissile material.

This proposed rule would revise the introductory paragraph by replacing (f) with
(g), paragraph (a) by adding new subparagraphs (1) and (2), paragraph (d) by replacing
“of up to” with “not exceeding, and add paragraph (g), which is a new provision for
exclusive use of transportation packages.

Section 71.17, Exemption from classification as fissile material.
This proposed rule would revise paragraph (e) to change the design approval
date for Type B or fissile material packages from April 1, 1996, to the effective date of
the final rule.

Section 71.19, Previously approved package.
This proposed rule would revise paragraph (a) to include existing CoCs that have
a “-96” in their package identification number, redesignate paragraphs (c) and (d) as
paragraphs (d) and (e), revise newly redesignated paragraph (e) to include those CoCs
that have a suffix “-96” in their identification numbers, and add new paragraph (c), to add
transitional arrangements on existing CoCs that have a “-96” in their package
identification number.

Section 71.22, General license: Fissile material.
This proposed rule would revise paragraph (a) to replace “subparts E and F of
this part” with “§§ 71.55 and 71.59” and to remove the limitation to a Type A quantity of
radioactive material in a Type A package to allow shipment of material under the general
licenses in §§ 71.22 and 71.23 in a Type B package, paragraph (c) to remove (c)(1) and
redesignate paragraph (c)(2) as new paragraph (c), paragraphs (e)(3) through (5) to limit
the 233U to less than one percent of the mass of 235U, similar to the provision limiting
plutonium in § 71.22(e)(5)(ii), and add new paragraphs (f) through (h) to ensure that
each licensee will comply with § 71.17(c) for shipments made using the respective
general license and that any Type B package used under the respective general license

approved by the NRC before the effective date of the final rule is subject to the
transitional arrangements in § 71.19.

Section 71.23, General license: Plutonium-beryllium special form material.
This proposed rule would revise paragraphs (a) and (c), and add paragraphs (f)
through (h) to clarify that only special form sealed sources, not just sealed sources may
be delivered to a carrier for transport using the general license in § 71.23.

Section 71.31, Contents of application.
This proposed rule would revise paragraph (a) to add a maintenance program
description, as required by § 71.35 among the contents of application.

Section 71.35, Package evaluation.
This proposed rule would revise paragraph (b) to delete “and” paragraph (c) to
add “; and” and add new paragraph (d) to specify maintenance program requirements.

Section 71.43, General standards for all packages.
This proposed rule would revise paragraph (d) to specifically include the
evaluation of the effects of aging, and to specify that degradation evaluations will be
managed by the maintenance program in accordance with § 71.35(d), and add new
paragraph (i) to specify that each system designed to contain liquids has adequate
ullage during evaluation of the tests and conditions for normal conditions of transport
and hypothetical accident conditions specified in §§ 71.71 and 71.73.

Section 71.55, General requirements for fissile material packages.
This proposed rule would revise paragraph (g)(1) to require that there is no
contact between the cylinder plug and any other part of the packaging, other than at its
original attachment point and that the cylinder plug remains leak tight, as NRC requires

for the cylinder valve.

Section 71.71, Normal conditions of transport.
This proposed rule would change the unit of measure in the table in paragraph
(c)(1) to change the unit of measure for the values of insolation used for the heat test for
normal conditions of transport from “(g cal/cm2)” to “(W/m2)”.

Section 71.73, Hypothetical accident conditions.
This proposed rule would revise paragraph (b) to add insolation to the initial
conditions for the tests for hypothetical accident conditions.

Section 71.77, Qualification of LSA-III Material.
This proposed rule would remove and reserve § 71.77 and make conforming
changes to §§ 71.4 and 71.100.

Section 71.95, Reports.
This proposed rule would remove paragraph (a)(3) as it is duplicative to text in
paragraph (b).

Section 71.97, Advance notification of shipment of irradiated reactor fuel and
nuclear waste.
This proposed rule would revise the section title, the introductory text of
paragraph (b), and paragraphs (d) and (f)(1) to remove references to irradiated reactor
fuel to correct a duplicative advance notification reporting requirement in § 71.97 with
those in §§ 73.35 and 73.37.

Section 71.100, Criminal penalties.

This proposed rule would revise paragraph (b) to remove the leaching test
requirement as a conforming change to § 71.77.

Section 71.106, Changes to quality assurance program.
This proposed rule would revise the introductory text of paragraph (b) to clarify
that a biennial report must be submitted to the NRC even if no changes are made to the
QAP during the reporting period.

Appendix A to Part 71 —Determination of A1 and A2
This proposed rule would revise Tables A-1 and A-2 in paragraph V.b. to add
seven radionuclides and correct the specific activity of natural rubidium.

VI.

Regulatory Flexibility Certification

Under the Regulatory Flexibility Act (5 U.S.C. 605(b)), the NRC certifies that this
proposed rule will not, if issued, have a significant economic impact on a substantial
number of small entities. This proposed rule affects a number of “small entities” as
defined by the Regulatory Flexibility Act or the size standards established by the NRC
(§ 2.810). However, as indicated in the regulatory analysis, these amendments do not
have a significant economic impact on the affected small entities.

VII.

Regulatory Analysis

The NRC has prepared a regulatory analysis on this proposed rule. The analysis
examines the costs and benefits of the alternatives considered by the NRC and includes
consideration of the costs and benefits of updating guidance. The NRC requests public
comment on the regulatory analysis. The regulatory analysis is available as indicated in
the “Availability of Documents” section of this document. Comments on the regulatory

analysis may be submitted to the NRC as indicated under the ADDRESSES section of
this document.

VIII.

Backfitting and Issue Finality

The NRC has determined that backfitting (§ 50.109, § 70.76, § 72.62, or § 76.76)
and the issue finality provisions in 10 CFR part 52 do not apply to this proposed rule
because it would not involve any provisions that would impose backfits as defined in 10
CFR chapter I or affect the issue finality of any approval issued under 10 CFR part 52.
Some licensees that are within the scope of the backfit rule (e.g., a power reactor or a
fuel fabrication facility) transport radioactive material from their own facilities. Those
backfitting and issue finality provisions apply to activities directly regulated under those
parts, and do not apply to activities regulated under other parts that do not include
backfitting or issue finality provisions. The exception to this general principle is where
the activity regulated under other parts that do not include backfitting or issue finality
provisions is an inextricable part of the regulated activity within the scope of backfitting
or issue finality. Preparing packages for transport is not an inextricable part of the
procedures or organization required to design, construct or operate a facility as licensed
under 10 CFR part 50, 52, 70, 72, or 76; rather, it is a separate activity that these
licensees may choose to undertake. The scope of this proposed rule does not include
any changes to any of those facilities or plants’ activities for which the backfit rule
applies.
The NRC’s determination on this matter is in accordance with Management
Directive 8.4, “Management of Backfitting, Forward Fitting, Issue Finality, and
Information Requests,” and its associated guidance in NUREG-1409, “Backfitting
Guidelines.”

IX.

Cumulative Effects of Regulation

The NRC seeks to minimize any potential negative consequences resulting from
the cumulative effects of regulation (CER). The CER describes the challenges that
licensees, or other impacted entities such as State partners, may face while
implementing new regulatory positions, programs, or requirements (e.g., rules, generic
letters, backfits, inspections). The CER is an organizational effectiveness challenge that
may result from a licensee or impacted entity implementing a number of complex
regulatory actions, programs, or requirements within limited available resources.
To better understand the potential CER implications incurred due to this
proposed rule, the NRC is requesting comment on the following questions. Responding
to these questions is voluntary, and the NRC will respond to any comments received in
the final rule.
1. In light of any current or projected CER challenges, does the proposed rule’s
effective date provide sufficient time to implement the new proposed requirements,
including changes to programs and procedures?
2. If current or projected CER challenges exist, what should be done to address
this situation? For example, if more time is required for implementation of the new
requirements, what period of time is sufficient?
3. Do other regulatory actions (from the NRC or other agency) influence the
implementation of the proposed rule’s requirements?
4. Are there unintended consequences? Does the proposed rule create
conditions that would be contrary to the proposed rule’s purpose and objectives? If so,
what are the unintended consequences, and how should they be addressed?
5. Please comment on the NRC’s cost and benefit estimates in the regulatory
analysis that supports this proposed rule.

X.

Plain Writing

The Plain Writing Act of 2010 (Pub. L. 111-274) requires Federal agencies to
write documents in a clear, concise, and well-organized manner. The NRC has written
this document to be consistent with the Plain Writing Act as well as the Presidential
Memorandum, “Plain Language in Government Writing,” published June 10, 1998 (63
FR 31885). The NRC requests comment on this document with respect to the clarity
and effectiveness of the language used.

XI.

Environmental Assessment and Proposed Finding of No Significant
Environmental Impact

The Commission has preliminarily 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, if adopted, would not be a major Federal action significantly
affecting the quality of the human environment, and an environmental impact statement
is not required. The basis of this determination is as follows: The amendments would
change the requirements for packaging and transportation of radioactive material. The
amendments would make changes to harmonize the NRC’s regulations with the 2018
Edition of the IAEA’s transport standards (SSR-6) and with that of the DOT’s regulations
under 49 CFR and include NRC-initiated changes. The environmental impacts arising
from the changes have been evaluated and would not involve any significant
environmental impact. This includes consideration of direct, indirect, and cumulative
impacts. Other amendments are procedural in nature and would have no significant
impact on the environment.
The preliminary determination of this environmental assessment is that there will
be no significant effect on the quality of the human environment from this action. Public
stakeholders should note, however, that comments on any aspect of this environmental
assessment may be submitted to the NRC as indicated under the ADDRESSES caption.

The environmental assessment is available as indicated under the “Availability of
Documents” section of this document.
The NRC has sent a copy of the environmental assessment and this proposed
rule to every State Liaison Officer and has requested comments.

XII.

Paperwork Reduction Act

This proposed rule contains new or amended information collection requirements
that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq). This
proposed rule has been submitted to the Office of Management and Budget (OMB) for
review and approval of the information collection requirements.
Type of submission, new or revision: Revision.
The title of the information collection: Harmonization of Transportation Safety
Requirements with IAEA Standards.
The form number if applicable: Not applicable.
How often the collection is required: Applications for changes reducing
commitments to the NRC on quality assurance programs and for package approval are
submitted on occasion. Quality assurance program reporting on changes determined
not to reduce commitments, or reporting of no changes made, is done every 24 months.
Reporting packaging issues or instances in which the conditions in a CoC are not
followed occur infrequently.
Who will be required or asked to report: General or specific licensees who use a
package, certificate holders and applicants for a new or amended CoC.
An estimate of the number of annual responses: 7.5.
The estimated number of annual respondents: 6.5.
An estimate of the total number of hours needed annually to complete the
requirement or request: 1376.7 hours (an increase of 1052.5 hours reporting + an
increase of 322.7 third party disclosure hours and 1.5 hours recordkeeping).

Abstract: The NRC, in consultation with the DOT, is proposing to amend its
regulations for the packaging and transportation of radioactive material. The
Commission has historically been consistent in its support of harmonizing the NRC
transportation regulations with the IAEA’s standards. These amendments would make
the NRC regulations conform to the recent revisions to the IAEA standards for the
international transportation of radioactive material and maintain consistency with the
DOT regulations. These changes are necessary to maintain a consistent regulatory
framework for the packaging and transportation of radioactive material. The NRC is also
proposing to amend these regulations to include administrative, editorial, or clarifying
changes, including changes to certain Agreement State compatibility category
designations.
The NRC is seeking public comment on the potential impact of the information
collections contained in this proposed rule and on the following issues:
1. Is the proposed information collection necessary for the proper performance of
the functions of the NRC, including whether the information will have practical utility?
2. Is the estimate of burden of the proposed information collection accurate?
3. Is there a way to enhance the quality, utility, and clarity of the information to be
collected?
4. How can the burden of the proposed information collection on respondents be
minimized, including the use of automated collection techniques or other forms of
information technology?
A copy of the OMB clearance package is available in ADAMS under Accession
No. ML20101F920. You may obtain information and comment submissions related to
the OMB clearance package by searching on https://www.regulations.gov under Docket
ID NRC-2016-0179.
You may submit comments on any aspect of these proposed information
collection(s), including suggestions for reducing the burden and on the above issues, by
the following methods:

ï‚·

Federal Rulemaking Website: Go to https://www.regulations.gov and

search for Docket ID NRC-2016-0179.
ï‚·

Mail comments to: FOIA, Library, and Information Collections Branch T6-

A10M, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by email
to Infocollects.Resource@nrc.gov.
ï‚·

Submit to OMB Directly: Written comments and recommendations for the

proposed information collection should be sent within 60 days of publication of this
document to https://www.reginfo.gov/public/do/PRAMain. Find this particular information
collection by selecting "Currently Under Review - Open for Public Comments" or by
using the search function.
Comments on the information collections will be publicly available in ADAMS and
on Reginfo.gov. Submit comments by [INSERT DATE 60 DAYS AFTER
PUBLICATION IN THE FEDERAL REGISTER]. Comments received after this date will
be considered if it is practical to do so, but the NRC is able to ensure consideration only
for comments received on or before this date.

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

XIII.

Criminal Penalties

For the purposes of Section 223 of the Atomic Energy Act of 1954, as amended
(AEA), the NRC is issuing this proposed rule that would amend 10 CFR part 71 under
one or more of Sections 161b, 161i, or 161o of the AEA. Willful violations of the rule
would be subject to criminal enforcement. With the following exception, none of the

proposed amendments would change the manner in which criminal penalties would be
assessed or enforced.
Criminal penalties as they apply to regulations in 10 CFR part 71 are discussed
in § 71.100. One of the actions within the scope of this rulemaking, Issue 6, Deletion of
the Low Specific Activity-III Leaching Test, proposes to remove the content of § 71.77
and replace the section heading with “RESERVED.” This change would impact
§ 71.100(b), because § 71.77 would be removed from that paragraph as the leaching
test would no longer be required.

XIV.

Coordination with NRC Agreement States

The NRC has coordinated with the Agreement States throughout the
development of this proposed rule. Agreement State representatives have served on
the rulemaking working group that developed this proposed rule and on the Standing
Committee on Compatibility for the rulemaking. The NRC also provided a preliminary
draft of the proposed rule to the Agreement States for review.

XV.

Compatibility of Agreement State Regulations

Under the “Agreement State Program Policy Statement” approved by the
Commission on October 2, 2017 and published in the Federal Register on October 18,
2017 (82 FR 48535), NRC program elements (including regulations) are placed into
compatibility categories A, B, C, D, NRC, or adequacy category Health and Safety
(H&S). Compatibility Category A program elements are those program elements that
are basic radiation protection standards and scientific terms and definitions that are
necessary to understand radiation protection concepts. An Agreement State should
adopt Category A program elements in an essentially identical manner in order to
provide uniformity in the regulation of agreement material on a nationwide basis.

Compatibility Category B program elements are those program elements that apply to
activities that have direct and significant effects in multiple jurisdictions. An Agreement
State should adopt Category B program elements in an essentially identical manner.
Compatibility Category C program elements are those program elements that do not
meet the criteria of Category A or B but do contain the essential objectives that an
Agreement State should adopt to avoid conflict, duplication, gaps, or other conditions
that would jeopardize an orderly pattern in the regulation of agreement material on a
national basis. An Agreement State should adopt the essential objectives of the
Category C program elements. Compatibility Category D program elements are those
program elements that do not meet any of the criteria of Category A, B, or C and,
therefore, do not need to be adopted by Agreement States for purposes of compatibility.
Compatibility Category NRC program elements are those program elements that
address areas of regulation that cannot be relinquished to the Agreement States under
the Atomic Energy Act of 1954, as amended, or provisions of title 10 of the Code of
Federal Regulations. These program elements should not be adopted by the Agreement
States. Adequacy category H&S program elements are program elements that are
required because of a particular health and safety role in the regulation of agreement
material within the State and should be adopted in a manner that embodies the essential
objectives of the NRC program. A bracketed compatibility category (e.g., [B]) means
that the provision may have been adopted elsewhere in the Agreement State’s
regulations and does not need to be adopted again.
As discussed in Section III of this document, Issue 15.4, the regulations that
contain QAP requirements (e.g., §§ 71.109, 71.111, 71.113, 71.115, 71.117, 71.119,
71.121, 71.123, and 71.125) are currently designated as Compatibility Category NRC
and cannot be adopted by the Agreement States. Since a proper QAP review cannot be
completed without addressing many of these criteria, Agreement States would need to
adopt compatible regulations to require licensees that use NRC-approved Type B
packages for shipping, other than for industrial radiography, or that ship using the

general license in § 71.21, § 71.22 or § 71.23, to follow these QAP criteria. Additionally,
since only a few Agreement States have applicable licensees that perform shipments of
Type B quantities of radioactive materials, other than for industrial radiography
operations (which are covered under § 34.31), or that ship using the general license in
§ 71.21, § 71.22, or § 71.23, all QAP-related requirements, including those mentioned
previously and others referenced below in the table, would be re-designated as a
Compatibility Category B. This re-designation would require those Agreement States
with applicable licensees to have essentially identical regulations. For those Agreement
States that do not have applicable licensees, these regulations will remain designated as
Compatibility Category D and, hence, do not have to be adopted for purposes of
compatibility.
The changes in this proposed rule, discussed in Section III of this document,
would be a matter of compatibility between the NRC and the Agreement States, thereby
providing consistency among Agreement State and NRC requirements. Regulations that
are a part of this rulemaking but remain the same compatibility category designation are
included in the table for completeness. The compatibility categories are designated in
the following table.

Section

Change

Subject

71.0(d)(1)
71.4

Revised
New

71.4

Revised

Purpose and Scope
Definition:
Radiation Level
Definition:
Low Specific Activity
(LSA) material

Compatibility
Existing

New

D
—

D
[A]

[B]

[B]

[B]

[B]

[B]

[B]

[B]

[B]

[Deletion of Low Specific
Activity-III Leaching Test]

71.4

Revised

71.4

Revised

71.15(a) and (d)

Revised

Definition:
Special form
radioactive material
Definition:
Surface Contaminated
Object (SCO)
Exemption from
classification as fissile
material

71.15(g)

New

-

[B]

B

B

NRC

NRC

[B]

[B]



[B]

[B]

[B]



[B]

Revised

Exemption from
classification as fissile
material
General license: NRCapproved package.
Previously approved
package
General license: Fissile
material
General license: Fissile
material
General license:
Plutonium-beryllium
special form material
General license:
Plutonium-beryllium
special form material
Contents of application

71.17(e)

Revised

71.19

Revised

71.22(a), (c), and (e)(3)
through (5)
71.22(f) through (h)

Revised

71.23(a) and (c)

Revised

71.23(f) through (h)

New

71.31(a)

NRC

NRC

71.35(b) and (c)

Revised

Package evaluation

NRC

NRC

71.35(d)

New

Package evaluation

--

NRC

71.43(d)

Revised

NRC

NRC

71.43(i)

New



NRC

71.55(g)

Revised

NRC

NRC

71.71(c)(1)

Revised

NRC

NRC

71.73(b)

Revised

NRC

NRC

71.77

Removed

NRC



71.95

Revised
compatibility
category
Removed
Revised

General standards for
all packages
General standards for
all packages
General requirements
for fissile material
packages
Normal conditions of
transport
Hypothetical accident
conditions
Qualification of LSA-III
Material
Reports

D

C**

Reports
Advance notification of
shipment of irradiated
reactor fuel and
nuclear waste
Criminal penalties
Quality assurance
requirements

D
B

*

D
C***

D
B***

Quality assurance
requirements

C***

B**

Quality assurance
organization

C***

B**

Quality assurance
organization

D

B**

71.95(a)(3)
71.97

71.100
71.101(b)
71.101(c)(1)
71.103
(a) and (b)
71.103
(c), (d), (e) and (f)

New

Revised
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category

B

71.105
71.106
71.109
71.111
71.113
71.115
71.117
71.119
71.121
71.123
71.125
71.127
71.129
71.131
71.133
71.135
71.137
Table A-1 in Appendix A
to 10 CFR Part 71

Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised
compatibility
category
Revised

Quality assurance
program

C

B**

Changes to quality
assurance program

C

B**

Procurement document
control

NRC

B**

Instructions,
procedures and
drawings
Document control

NRC

B**

NRC

B**

Control of purchased
material, equipment,
and services
Identification and
control of materials,
parts and components
Control of special
processes

NRC

B**

NRC

B**

NRC

B**

Internal inspection

NRC

B**

Test control

NRC

B**

Control of measuring
and test equipment

NRC

B**

Handling, storage, and
shipping control

[C]

B**

Inspection, test, and
operating status

[C]

B**

Nonconforming
materials, parts, or
components
Corrective action

[C]

B**

C

B**

Quality assurance
records

C***

C**

Audits

C

C**

A1 and A2 Values for
Radionuclides

[B]

[B]

Table A-2 in Appendix A
to 10 CFR Part 71

Revised

Exempt Material
Activity Concentrations
and Exempt
Consignment Activity
Limits for
Radionuclides

[B]

[B]

* Denotes regulations that are designated Compatibility Category D but which will be removed from the
regulations as a result of these proposed amendments. Agreement States that have an equivalent regulation
should remove these provisions from their regulations when the regulations become final.
** B/C (as designated) – for Agreement States that have licensees that use Type B approved packages for
shipping, other than for industrial radiography, or have licensees that ship using the general license in
§ 71.21, § 71.22, or § 71.23, these regulations are required for compatibility purposes.
D-for States that do not have licensees that use Type B approved packages for shipping, other than for
industrial radiography, these regulations are not required for compatibility purposes.
***10 CFR 71.101(g) indicates that QA programs for industrial radiography Type B package users are
covered by § 34.31(b). It also indicated that this section satisfies § 71.17(b) and therefore will satisfy those
sections referenced in this provision (§§ 71.101 through 71.137).

The NRC invites comment on the compatibility category designations in the
proposed rule and suggests that commenters refer to Handbook 5.9 of Management
Directive 5.9, “Adequacy and Compatibility of Program Elements for Agreement State
Programs,” for more information. The NRC notes that, like the rule text, the compatibility
category designations can change between the proposed rule and final rule on the basis
of comments received and Commission decisions regarding the final rule. The NRC
encourages anyone interested in commenting on the compatibility category designations
to do so during the comment period.

XVI.

Voluntary Consensus Standards

The National Technology Transfer and Advancement Act (NTTAA) of 1995, Pub.
L. 104-113, requires that Federal agencies use technical standards that are developed
or adopted by voluntary consensus standards bodies, unless the use of such a standard
is inconsistent with applicable law or otherwise impractical. In this proposed rule, the
NRC would revise regulations associated with packaging and transportation of
radioactive material in 10 CFR part 71 to conform NRC regulations to the recent
revisions to the IAEA standards for the international transportation of radioactive
material. While the rule harmonizes NRC requirements with IAEA Standard SSR-6, it
does not endorse SSR-6, and SSR-6 does not meet the criteria for being a voluntary

consensus standard under the NTTAA. The NRC is not aware of any voluntary
consensus standard that could be used. The NRC will consider using a voluntary
consensus standard if an appropriate standard is identified. If a voluntary consensus
standard is identified for consideration, the submittal should explain how the voluntary
consensus standard is comparable and why it should be used. This action does not
constitute the establishment of a standard that contains generally applicable
requirements.

XVII.

Availability of Guidance

The NRC is issuing for comment draft guidance, DG-7011, “Standard Format and
Content of Part 71 Applications for Approval of Packages for Radioactive Material,”
Revision 3 to Regulatory Guide 7.9, for the implementation of the requirements in this
proposed rule. The draft guidance identifies the information to be provided in an
application for package approval and establishes a uniform format for presenting that
information. The draft guidance is available in ADAMS under Accession No.
ML22223A085. You may obtain information and comment submissions related to the
draft guidance by searching on https://www.regulations.gov under Docket ID NRC-20160179. You may submit comments on the draft regulatory guidance by the methods
outlined in the ADDRESSES section of this document.
The NRC considered whether a revision of NUREG-1608, “Categorizing and
Transporting Low Specific Activity Materials and Surface Contaminated Objects,” was
warranted in association with this proposed rule. NUREG-1608, published jointly by the
NRC and the DOT in 1998, provides guidance to shippers of LSA material and SCO
regarding significant changes to both 10 CFR part 71 and 49 CFR that became effective
April 1, 1996. The NRC’s judgement is that NUREG-1608 serves the purpose for which
it was intended, which was to educate shippers about major changes to the regulations

in 1996, and that the minor changes to the LSA and SCO requirements in this proposed
rule do not warrant a revision to NUREG-1608.
The NRC also considered whether a revision of NUREG-1660, “U.S.-Specific
Schedules of Requirements for Transport of Specified Types of Radioactive Material
Consignments,” was warranted in association with this proposed rule. NUREG-1660,
published jointly by the NRC and the DOT in 1999, provides summaries of NRC, DOT,
and other regulations that shippers must meet, depending on the type of material being
shipped. NUREG-1660 is currently under revision to incorporate requirements issued in
both 10 CFR chapter I and 49 CFR chapter I since 1999. The NRC’s judgement is that
there are no changes being considered in this proposed rule that will affect the content
of the revised NUREG-1660.
The NRC considered whether a revision to NUREG-1886, “Joint Canada - United
States Guide for Approval of Type B(U) and Fissile Material Transportation Packages,”
is warranted in association with this rulemaking. NUREG-1886, published jointly with the
DOT and the Canadian Nuclear Safety Commission (CNSC) in 2009, provides a
standard format and content of an application for approval of Type B(U) and fissile
material packages to demonstrate the ability of the given package to meet both United
States (NRC and DOT regulations) and Canadian regulations. The NRC, the DOT, and
the CNSC recently started discussions to update NUREG-1886, which will be a
multiyear effort. When NUREG-1886 is updated, the NRC will ensure that it is
consistent with the final version of DG-7011 and its associated Regulatory Guide 7.9.
The NRC considered whether a revision to NUREG-2216, “Standard Review
Plan for Transportation Packages for Spent Fuel and Radioactive Material,” is warranted
in association with this proposed rule. NUREG-2216, which was recently issued,
provides guidance to the NRC staff for reviewing an application for package approval
issued under 10 CFR part 71. There are no changes being considered in this proposed
rule that would significantly affect the content of NUREG-2216. The NRC will first obtain

experience using NUREG-2216 to evaluate whether there are more significant changes
needed before making the relatively minor changes associated with this proposed rule.

XVIII.

Public Meeting

The NRC will conduct a public meeting on this proposed rule to describe it to the
public and to facilitate the development of public comments. The NRC will publish a
notice of the location, time, and agenda of the meeting on Regulations.gov and on the
NRC’s public meeting Web site at least 10 calendar days before the meeting.
Stakeholders should monitor the NRC’s public meeting Web site for information about
the public meeting at: https://www.nrc.gov/public-involve/public-meetings/index.cfm.

XIX.

Availability of Documents

The documents identified in the following table are available to interested
persons through one or more of the following methods, as indicated.
DOCUMENT
Rulemaking Documents and References
SECY-20-0102 for this proposed rule
Federal Register notice for this proposed rule
Regulatory Analysis for this proposed rule
Environmental Assessment for this proposed rule
OMB supporting statement for this proposed rule
Draft regulatory basis document for this rulemaking,
dated March 2019
Federal Register notification for draft regulatory
basis, dated April 12, 2019
Draft regulatory basis comment submission #1
Draft regulatory basis comment submission #2
Draft regulatory basis comment submission #3
Draft regulatory basis comment submission #4
Draft regulatory basis comment submission #5
Draft regulatory basis comment submission #6
Draft regulatory basis comment submission #7
NRC final rule amending packaging and
transportation of radioactive material regulations,
dated June 12, 2015

ADAMS ACCESSION NO. /
WEB LINK / FEDERAL
REGISTER CITATION
ML20101F921
ML22209A035
ML22209A039
ML22209A045
ML22209A052
ML18262A185
84 FR 14898
ML19106A347
ML19113A064
ML19143A311
ML19143A312
ML19148A147
ML19149A474
ML19150A140

80 FR 33988

DOT final rule amending packaging and
transportation of radioactive material regulations,
dated July 11, 2014
NRC final rule harmonizing its regulations with the
1996 edition of IAEA Safety Series No. 6, dated
January 26, 2004
NRC proposed rule harmonizing its regulations with
the 1996 edition of IAEA Safety Series No. 6, dated
April 30, 2002
NRC final rule harmonizing its regulations with the
1985 edition of IAEA Safety Series No. 6, dated
September 28, 1995
NRC/DOT Memorandum of Understanding, dated
July 2, 1979
SECY-16-0093, “Rulemaking Plan for Revisions to
Transportation Safety Requirements and
Harmonization with International Atomic Energy
Agency Transportation Requirements,” dated July
28, 2016
Staff Requirements Memorandum SRM-SECY-160093, “Staff Requirements – SECY-16-0093 –
Rulemaking Plan for Revisions to Transportation
Safety Requirements and Harmonization with
International Atomic Energy Agency Transportation
Requirements,” dated August 19, 2016
Harmonization issues paper, “Issues Paper on
Potential Revisions to Transportation Safety
Requirements and Harmonization with International
Atomic Energy Agency Transportation
Requirements,” dated November 15, 2016
Federal Register notification for harmonization
issues paper, dated November 21, 2016
Issues paper public meeting summary, “Summary of
the December 5 and 6, 2016 Public Meeting on
Issues Paper on Revisions to Transportation Safety
Requirements and Harmonization with the
International Atomic Energy Agency Transportation
Requirements,” dated December 14, 2016
Draft Regulatory Guidance Document
Draft Regulatory Guide DG-7011, “Standard Format
and Content of Part 71 Applications for Approval of
Packages for Radioactive Material,” Revision 3 of
Regulatory Guide 7.9

79 FR 40589
69 FR 3697
67 FR 21390
60 FR 50248
44 FR 38690
ML16158A164

ML16235A182

ML16299A298 paper
ML16299A291 package

81 FR 83171
ML16343A661

ML22223A085

IAEA Transportation Safety Standards and Related References
SSR-6, “Regulations for the Safe Transport of
https://www.iaea.org/publicatio
Radioactive Material,” 2018 Edition
ns/12288/regulations-for-thesafe-transport-of-radioactivematerial
SSR-6, “Regulations for the Safe Transport of
https://www.iaea.org/publicatio
Radioactive Material,” 2012 Edition
ns/8851/regulations-for-thesafe-transport-of-radioactivematerial-2012-edition

TS-R-1, “Regulations for the Safe Transport of
Radioactive Material,” 2009 Edition
TS-R-1, “Regulations for the Safe Transport of
Radioactive Material,” 2005 Edition
TS-R-1, “Regulations for the Safe Transport of
Radioactive Material,” 1996 Edition
Safety Series No. 6, “Regulations for the Safe
Transport of Radioactive Material, 1985 Edition (As
Amended in 1990)”
Safety Series No. 6, “Regulations for the Safe
Transport of Radioactive Material,” 1985 Edition
Safety Series No. 6, “Regulations for the Safe
Transport of Radioactive Material,” 1973 Edition
Safety Series No. 6, “Regulations for the Safe
Transport of Radioactive Material,” 1967 Edition
Other International Standards References
ANSI N14.1-2012, “Nuclear Materials - Uranium
Hexafluoride — Packagings for Transport,” dated
December 3, 2012
ANSI N14.5-2014, “American National Standard for
Radioactive Materials — Leakage Tests on
Packages for Shipment,” dated June 19, 2014
International Organization for Standardization
7195:2005, “Nuclear Energy–Packaging of Uranium
Hexafluoride (UF6) for Transport,” dated September
2005
American National Standards Institute/American
Nuclear Society 8.1-2014 (Reaffirmed 2018),
“Nuclear Criticality Safety in Operations with
Fissionable Materials Outside Reactors,” American
Nuclear Society, La Grange Park, IL
Miscellaneous References
National Renewable Energy Laboratory Solar
Radiation Data
NRC letter to Agreement States, “Clarification of Title
10 of the Code of Federal Regulations, Part 71
Requirements Identified in Regulation Amendment
Tracking System Identification Number RATS ID:
2015-3 (STC-17-060),” dated August 15, 2017
Presidential Memorandum, “Plain Language in
Government Writing,” published June 10, 1998
Agreement State Program Policy Statement, dated
October 18, 2017

https://www.iaea.org/publicatio
ns/8005/regulations-for-thesafe-transport-of-radioactivematerial-2009-edition
https://www.iaea.org/publicatio
ns/7291/regulations-for-thesafe-transport-of-radioactivematerial-2005-edition
https://www.iaea.org/publicatio
ns/6056/regulations-for-thesafe-transport-of-radioactivematerial-1996-edition-revised
http://gnssn.iaea.org/Superse
ded%20Safety%20Standards/
Safety_Series_006_1990.pdf
https://gnssn.iaea.org/Supersede
d%20Safety%20Standards/Safet
y_Series_006_1985.pdf
https://gnssn.iaea.org/Supersede
d%20Safety%20Standards/Safet
y_Series_006_1973.pdf
https://gnssn.iaea.org/Supersede
d%20Safety%20Standards/Safet
y_Series_006_1967.pdf

https://webstore.ansi.org/stan
dards/pcc/ansin142012
https://webstore.ansi.org/stan
dards/pcc/ansin142014
https://www.iso.org/standard/3
1251.html
https://webstore.ansi.org/Standar
ds/ANSI/ANSIANS2014R2018

https://www.nrel.gov/gis/asset
s/images/solar-annual-ghi2018-usa-scale-01.jpg
ML17213A844

63 FR 31885
82 FR 48535

NRC Management Directive 5.9, Handbook 5.9,
“Adequacy and Compatibility of Program Elements
for Agreement State Programs,” dated April 26, 2018
NRC Management Directive 8.4, “Management of
Backfitting, Forward Fitting, Issue Finality, and
Information Requests,” dated September 20, 2019
ORNL/TM-2014/658, “Comparison of the
International and United States Domestic
Radioactive Material Transport Regulations,” dated
September 30, 2014
NUREG-1409, “Backfitting Guidelines,” Revision 1,
draft for public comment, dated March 2020
NUREG-1608, “Categorizing and Transporting Low
Specific Activity Materials and Surface Contaminated
Objects,” dated July 1998
NUREG-1660, “U.S.-Specific Schedules of
Requirements for Transport of Specified Types of
Radioactive Material Consignments,” dated January
1999
NUREG-1886, “Joint Canada — United States Guide
for Approval of Type B(U) and Fissile Material
Transportation Packages,” dated March 2009
NUREG-2216, “Standard Review Plan for
Transportation Packages for Spent Fuel and
Radioactive Material,” dated August 2020

ML18081A070
ML18093B087
https://rampac.energy.gov/doc
s/defaultsource/doeinfo/ORNL-TM2014-658.pdf
ML18109A498
ML15336A927
https://rampac.energy.gov/doc
s/defaultsource/nrcinfo/nureg_1660.pdf
ML090930197
ML20234A651

Throughout the development of this proposed rule, the NRC may post
documents related to it, including public comments, on the Federal rulemaking Web site
at https://www.regulations.gov under Docket ID NRC-2016-0179. In addition, the
Federal rulemaking website allows members of the public to receive alerts when
changes or additions occur in a docket folder. To subscribe: 1) navigate to the docket
folder (NRC-2016-0179); 2) click the “Subscribe” link; and 3) enter an email address and
click on the “Subscribe” link.

List of Subjects in 10 CFR Part 71

Criminal penalties, Hazardous materials transportation, Intergovernmental
relations, Nuclear materials, Packaging and containers, Penalties, Radioactive materials,
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, the NRC is proposing to adopt the following amendments to
10 CFR part 71:

PART 71 – PACKAGING AND TRANSPORTATION OF RADIOACTIVE MATERIAL

1. The authority citation for part 71 continues to read as follows:
Authority: Atomic Energy Act of 1954, secs. 53, 57, 62, 63, 81, 161, 182, 183,
223, 234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093, 2111, 2201, 2232, 2233, 2273,
2282, 2297f); Energy Reorganization Act of 1974, secs. 201, 202, 206, 211 (42 U.S.C.
5841, 5842, 5846, 5851); Nuclear Waste Policy Act of 1982, sec. 180 (42 U.S.C. 10175);
44 U.S.C. 3504 note.
Section 71.97 also issued under Sec. 301, Pub. L. 96-295, 94 Stat. 789 (42
U.S.C. 5841 note).
2. In § 71.0, revise paragraph (d)(1) to read as follows:
§ 71.0 Purpose and scope.
*

*

*

*

*

(d)(1) Exemptions from the requirement for license in § 71.3 are specified in
§ 71.14. The general license in § 71.21 does not require NRC package approval. The
general licenses in §§ 71.22 and 71.23 require NRC package approval if the quantities
exceed a Type A quantity. The general license in § 71.17 requires that an NRC
certificate of compliance or other package approval be issued for the package to be
used under this general license.
*

*

*

*

*

3. Amend § 71.4 by:
a. Revising the definitions for Low Specific Activity material and Special form
radioactive material;
b. Revising the introductory text and add paragraph (3) for Surface contaminated
object; and

c. Adding the definition Radiation level in alphabetical order.
The revisions and addition read as follows:

§ 71.4 Definitions.
*

*

*

*

*

Low Specific Activity (LSA) material means radioactive material with limited
specific activity which is nonfissile or is exempt under § 71.15, and which satisfies the
descriptions and limits set forth in the following section. Shielding materials surrounding
the LSA material may not be considered in determining the estimated average specific
activity of the package contents. The LSA material must be in one of three groups:
* * * * *
(3) LSA—III. Solids (e.g., consolidated wastes, activated materials), excluding
powders, in which:
(i) The radioactive material is distributed throughout a solid or a collection of solid
objects, or is essentially uniformly distributed in a solid compact binding agent (such as
concrete, bitumen, ceramic, etc.); and
(ii) [Reserved]
(iii) The estimated average specific activity of the solid, excluding any shielding
material, does not exceed 2 x 10–3A2/g.
*

*

*

*

*

Radiation level means the radiation dose equivalent rate expressed in
millisieverts per hour or mSv/h (millirems per hour or mrem/h).
*

*

*

*

*

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 § 71.75. A special form encapsulation
designed in accordance with the requirements of § 71.4 in effect from April 1, 1996, to
September 30, 2004, may continue to be used, provided that fabrication of the special
form encapsulation was successfully completed by [DATE ONE DAY PRIOR TO
EFFECTIVE DATE OF FINAL RULE]. A special form encapsulation designed in
accordance with the requirements of § 71.4 in effect from October 1, 2004, to [DATE
ONE DAY PRIOR TO EFFECTIVE DATE OF FINAL RULE] may continue to be used,
provided that fabrication of the special form encapsulation is successfully completed by
December 31, 2025. Any other special form encapsulation must meet the specifications
of this definition.
*

*

*

*

*

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 three groups with surface activity not
exceeding the following limits:
*

*

*

*

*

(3) SCO-III: A large solid object which, because of its size, cannot be
transported in a type of package described in 49 CFR 173.403 of the DOT regulations
and for which:
(i) All openings are sealed to prevent release of radioactive material during
conditions defined in 49 CFR 173.427(d);
(ii) The inside of the object is as dry as practicable;
(iii) The nonfixed contamination on the external surface does not exceed the
contamination limits specified in the DOT regulations in 49 CFR 173.443; and
(iv) The nonfixed contamination plus the fixed contamination on the
inaccessible surface averaged over 300 cm2 does not exceed 8 x 105 Bq/cm2 (20

microcuries/cm2) for beta and gamma emitters and low toxicity alpha emitters, or 8 x
104 Bq/cm2 (2 microcuries/cm2) for all other alpha emitters.
*

*

*

*

*

4. In § 71.15, revise the introductory text and paragraphs (a) and (d) and add
paragraph (g) to read as follows:

§ 71.15 Exemption from classification as fissile material.
Fissile material meeting the requirements of at least one of the paragraphs (a)
through (g) of this section are exempt from classification as fissile material and from the
fissile material package standards of §§ 71.55 and 71.59 but are subject to all other
requirements of this part, except as noted.
(a) Individual package containing:
(1) 2 grams or less fissile material, or
(2) 3.5 grams or less uranium-235, provided the uranium is enriched in uranium235 to a maximum of 5 percent by weight, and the total plutonium and uranium-233
content does not exceed 1 percent of the mass of uranium-235.
*

*

*

*

*

(d) Uranium enriched in uranium-235 to a maximum of 1 percent by weight, and
with total plutonium and uranium-233 content not exceeding 1 percent of the mass of
uranium-235, provided that the mass of any beryllium, graphite, and hydrogenous
material enriched in deuterium constitutes less than 5 percent of the uranium mass, and
that the fissile material is distributed homogeneously and does not form a lattice
arrangement within the package.
*

*

*

*

*

(g) Packages transported under exclusive use on a conveyance containing a
total of 140 grams or less fissile material.

5. In § 71.17, revise paragraph (e) to read as follows:
§ 71.17 General license: NRC-approved package.
*

*

*

*

*

(e) For a Type B or fissile material package, the design of which was approved
by NRC before [EFFECTIVE DATE OF FINAL RULE], the general license is subject to
the additional restrictions of § 71.19.

6. Amend § 71.19 by:
a. Revising paragraph (a);
b. Redesignating paragraphs (c) and (d) as paragraphs (d) and (e);
c. Adding new paragraph (c); and
d. Revising newly redesignated paragraph (e).
The revisions and addition read as follows:

§ 71.19 Previously approved package.
(a) A Type B(U) package, a Type B(M) package, or a fissile material package,
previously approved by the NRC but without the designation “-85” or “-96” in the
identification number of the NRC CoC, may be used under the general license of § 71.17
with the following additional conditions:
(1) Fabrication of the package is satisfactorily completed by April 1, 1999, as
demonstrated by application of its model number in accordance with § 71.85(c);
(2) 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; and

(3) Paragraph (a) of this section expires [DATE 8 YEARS AFTER EFFECTIVE
DATE OF THE FINAL RULE].
*

*

*

*

*

(c) A Type B(U) package, a Type B(M) package, or a fissile material package
previously approved by the NRC with the designation “-96” in the identification number of
the NRC CoC, may be used under the general license of § 71.17 with the following
additional conditions:
(1) Fabrication of the package must be satisfactorily completed by January 1,
2029, as demonstrated by application of its model number in accordance with
§ 71.85(c); and
(2) A package used for a shipment to a location outside the United States, after
December 31, 2025, is subject to multilateral approval, as defined in the DOT's
regulations at 49 CFR 173.403.
*

*

*

*

*

(e) NRC will revise the package identification number to designate previously
approved package designs that were designated as AF, B(U), B(M), B(U)F, B(M)F, B(U)85, B(U)F-85, B(M)-85, B(M)F-85, AF-85, B(U)-96, B(U)F-96, B(M)-96, B(M)F-96, or AF96 as appropriate, with the identification number suffix AF, B(U), B(M), B(U)F, B(M)F,
after receipt of an application demonstrating that the design meets the requirements of
this part.

7. In § 71.22, revise paragraphs (a), (c), and (e)(3) through (5) and add
paragraphs (f) through (h) to read as follows:

§ 71.22 General license: Fissile material.
(a) A general license is issued to any licensee of the Commission to transport
fissile material, or to deliver fissile material to a carrier for transport, if the material is
shipped in accordance with this section. The fissile material need not be contained in a

package which meets the standards of §§ 71.55 and 71.59. However, the material must
be contained in a Type A or Type B package, consistent with the quantity of radioactive
material in the package.
*

*

*

*

*

(c) The general license applies only when a package's contents contain less than
500 total grams of beryllium, graphite, or hydrogenous material enriched in deuterium.
*

*

*

*

*

(e) * * *
(3) The values of X, Y, and Z used in the CSI equation must be taken from Table
71-1 or 71-2, as appropriate based on criteria from § 71.22(e)(4) and (5).
(4) If Table 71-2 is used to obtain the value of X, then:
(i) The total mass of plutonium and uranium-233 must not exceed 1 percent of
the mass of uranium-235;
(ii) Values for the terms in the equation for uranium-233 and plutonium must be
assumed to be zero; and
(iii) The value of the uranium enrichment must be known and be less than the
enrichment value used from Table 71-2.
(5) Table 71-1 values for X, Y, and Z must be used to determine the CSI if:
(i) The total mass of plutonium and uranium-233 exceeds 1 percent of the mass
of uranium-235;
(ii) The uranium is of unknown uranium-235 enrichment or greater than 24 weight
percent enrichment; or

(iii) Substances having a moderating effectiveness (i.e., an average hydrogen
density greater than H2O) (e.g., certain hydrocarbon oils or plastics) are present in any
form, except as polyethylene used for packing or wrapping. *
*

*

*

*

*

*

*

(f) Each licensee using the general license under paragraph (a) of this section to
transport a Type B quantity of licensed material must use a package for which a license,
CoC, or other approval has been issued by the NRC, and must comply with the
provisions in § 71.17(c).
(g) For shipment of a Type B quantity of licensed material, this general license
applies only when the package approval authorizes use of the package under the
general license in § 71.17 or this general license.
(h) For a Type B package, the design of which was approved by NRC before
[EFFECTIVE DATE OF FINAL RULE], this general license is subject to the additional
restrictions of § 71.19.

8. In § 71.23, revise paragraph (a) and the introductory text of paragraph (c) and
add paragraphs (f) through (h) to read as follows:

§ 71.23 General license: Plutonium-beryllium special form material.
(a) A general license is issued to any licensee of the Commission to transport
fissile material in the form of plutonium-beryllium (Pu-Be) special form sources, or to
deliver Pu-Be special form sources to a carrier for transport, if the material is shipped in
accordance with this section. This material need not be contained in a package which
meets the standards of §§ 71.55 and 71.59. However, the fissile material must be
contained in a Type A or Type B package, consistent with the quantity of radioactive
material in the package.
*

*

*

*

*

(c) The general license applies only when a package's contents contain less than
1000 grams of plutonium, provided that plutonium-239, plutonium-241, or any
combination of these radionuclides, constitutes less than 240 grams of the total quantity
of plutonium in the package.
*

*

*

*

*

(f) Each licensee using the general license under paragraph (a) of this section to
transport a Type B quantity of licensed material must use a package for which a license,
CoC, or other approval has been issued by the NRC, and must comply with the
provisions in § 71.17(c).
(g) For shipment of a Type B quantity of licensed material, this general license
applies only when the package approval authorizes use of the package under the
general license in § 71.17 or this general license.
(h) For a Type B package, the design of which was approved by NRC before
[EFFECTIVE DATE OF FINAL RULE], this general license is subject to the additional
restrictions of § 71.19.

9. In § 71.31, revise paragraph (a) to read as follows:

§ 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 § 71.33;
(2) A package evaluation as required by § 71.35;
(3) A maintenance program description, as required by § 71.35; and
(4) A quality assurance program description, as required by § 71.37, or a
reference to a previously approved quality assurance program.
*

*

*

*

*

10. In § 71.35, revise paragraphs (b) and (c) and add paragraph (d) to read as
follows:
§ 71.35 Package evaluation.
*

*

*

*

*

(b) For a fissile material package, the allowable number of packages that may be
transported in the same vehicle in accordance with § 71.59;
(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; and
(d) A maintenance program to assure that the packaging will perform as intended
throughout its time in service. The maintenance program must include periodic testing
requirements, inspections, and replacement criteria and schedules for replacement and
repairs of components on an as-needed basis.

11. In § 71.43, revise paragraph (d) and add paragraph (i) to read as follows:

§ 71.43 General standards for all packages.
*

*

*

*

*

(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. The effects of the aging mechanisms and the behavior of
materials under irradiation must be evaluated on package components to show that their
performance is not significantly degraded or that degradation will be managed by the
maintenance program in accordance with § 71.35(d).
*

*

*

*

*

(i) Each system designed for holding liquids must be designed, constructed, and
prepared for shipment so that under the tests specified in §§ 71.71 and 71.73, there
would be adequate space to accommodate variations in temperature of the liquid,
dynamic effects, and filling dynamics.

12. In § 71.55, revise paragraph (g)(1) to read as follows:

§ 71.55 General requirements for fissile material packages.
*

*

*

*

*

(g) * * *
(1) Following the tests specified in § 71.73 (“Hypothetical accident
conditions”), there is no physical contact between the valve body or the plug and
any other component of the packaging, other than at its original point of
attachment, and the valve and plug remain leak tight;
*

*

*

*

*

13. In § 71.71, in the table in paragraph (c)(1), revise the heading of the second
column to read as follows:
§ 71.71 Normal conditions of transport.
*****
(c) * * *
(1) * * *
Insolation Data
***

Total insolation for a 12-hour period
(W/m2)

*******

*****

14. In § 71.73, revise paragraph (b) to read as follows:

§ 71.73 Hypothetical accident conditions.
*

*

*

*

*

(b) Test conditions. Except for the water immersion test, the following
conditions shall apply before and after the tests:
(1) The ambient air temperature shall remain constant at that value
between -29 °C (-20 °F) and +38 °C (+100 °F) which is most unfavorable for the
feature under consideration;
(2) The insolation shall be that value between 0 and the maximum value
listed in the Insolation Data Table in § 71.71(c)(1), which is most unfavorable for
the feature under consideration; and
(3) 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.
*

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*

*

*

§ 71.77 [Removed and Reserved]
15. Remove and reserve § 71.77.

§ 71.95 [Amended]
16. In § 71.95, remove paragraph (a)(3).

§ 71.97 [Amended]
17. In § 71.97:

a. In the section heading, remove the phrase “irradiated reactor fuel and”;
b. In paragraph (b) introductory text, remove the word “also”;
c. In paragraph (d) introductory text and paragraphs (d)(1) and (2), remove the
phrase “irradiated reactor fuel or”; and
d. In paragraph (f)(1), remove the phrase “an irradiated reactor fuel or” and add
in its place the word “a”.

§ 71.100 [Amended]
18. In § 71.100(b), remove the reference “71.77,”.

19. In § 71.106, revise the introductory text of paragraph (b) to read as follows:

§ 71.106 Changes to quality assurance program.
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*

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*

*

(b) Each quality assurance program approval holder may change a previously
approved quality assurance program without prior NRC approval, if the change does not
reduce the commitments in the quality assurance program previously approved by the
NRC. Changes to the quality assurance program that do not reduce the commitments
shall be submitted to the NRC every 24 months, in accordance with § 71.1(a). If no
changes were made to the quality assurance program this information shall also be
submitted to the NRC every 24 months, in accordance with § 71.1(a). In addition to
quality assurance program changes involving administrative improvements and
clarifications, spelling corrections, and non-substantive changes to punctuation or
editorial items, the following changes are not considered reductions in commitment:
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20. In appendix A to part 71, in paragraph V.b.:

a. In Table A–1, add the entries for Ba-135m, Ge-69, Ir-193m, Ni-57, Sr-83, Tb149, and Tb-161 in alphanumeric order and revise the entries for Ni-59, Rb(nat), and Tb157; and
b. In Table A–2, add the entries for Ba-135m, Ge-69, Ir-193m, Ni-57, Sr-83, Tb149, and Tb-161 in alphanumeric order and revise the entries for Ni-59, Tb-157, Th(nat),
and U(nat).
The additions and revisions read as follows:
Appendix A to Part 71—Determination of A1 and A2
*

*

*

*

*

V.b. * * *
TABLE A-1—A1 AND A2 VALUES FOR RADIONUCLIDES
Element
and
Symbol of atomic
radionuclide number A1 (TBq)
*

*

Ba-135m
*

*

Ge-69
*

*

Ir-193m

Specific activity

A1 (Ci)b

A2 (TBq)

*

*

*

2.0 × 101

5.4 × 102

6.0 × 10−1

*

*

*

1.0 × 100

2.7 × 101

1.0 × 100

*

*

*

4.0 × 101

1.1 × 103

4.0 × 100

A2 (Ci)b

1.6 × 101

2.7 × 101

1.1 × 102

(TBq/g)

(Ci/g)

*

*

3.0 x 104

8.1 x 105

*

*

4.3 x 104

1.2 x 106

*

*

2.4 x 103

6.4 x 104

*

*
1.5 x 106

*

*

*

*

*

Ni-57

Nickel
(28)

6.0 ×
10−1

1.6 × 101

5.0 × 10−1

1.4 × 101

5.7 x 104

Unlimited

Unlimited

Unlimited

Unlimited

3.0 X 10-3 8.0 X 10-2

*

*

*

*

Unlimited

Unlimited

Unlimited

Unlimited 6.7 × 10-10 1.8 × 10-8

*

*

*

*

*

1.0 × 100

2.7 × 101

1.0 × 100

4.3 x 104

1.2 x 106

*

*

1.9 x 105

5.1 x 106

Ni-59
*

*

Rb(nat)
*

*

Sr-83
*

*

*

*

*

Tb-149

Terbiu
m (65)

8.0 × 10-1

2.2 × 101

8.0 × 10-1

2.7 × 101

2.2 × 101

*

Tb-157
*

1.1 X 103

4.0 X 101

*

*

*

3.0 × 101

8.1 × 102

7.0 × 10−1

*

*

*

*

Tb-161
*

4.0 X 101

*

1.1 X 103

1.9 × 101

5.6 X 10-1 1.5 X 101
*

*

4.3 x 103

1.2 x 105

*

*

TABLE A-2—EXEMPT MATERIAL ACTIVITY CONCENTRATIONS AND
EXEMPT CONSIGNMENT ACTIVITY LIMITS FOR RADIONUCLIDES

Symbol of
radionuclide

Element
and
atomic
number

*

*

*

Ba-135m
*

*

*

Ge-69
*

*

*

Ir-193m
*

*

*

Ni-57

Nickel (28)

Ni-59
*

*

*

Sr-83
*

*

Tb-149

Terbium
(65)

*

Tb-157
*

*

*

Tb-161
*

*

*

Th(nat) (b),
(c)
*

*

*

U(nat) (b), (c)
*

*

*

Activity
concentratio
n for exempt
material
(Bq/g)

Activity
concentratio
n for exempt
material
(Ci/g)

Activity limit
for exempt
consignment
(Bq)

Activity limit
for exempt
consignment
(Ci)

*

*

*

*

1.0 × 102

2.7 × 10−9

1.0 × 106

2.7 × 10−5

*

*

*

*

1.0 × 101

2.7 × 10−10

1.0 × 106

2.7 × 10−5

*

*

*

*

1.0 × 104

2.7 × 10−7

1.0 × 107

2.7 × 10−4

*

*

*

*

1.0 × 101

2.7 × 10−10

1.0 × 106

2.7 × 10−5

1.0 × 104

2.7 × 10−7

1.0 × 108

2.7 × 10−3

*

*

*

*

1.0 × 101

2.7 × 10−10

1.0 × 106

2.7 × 10−5

*

*

*

*

1.0 × 101

2.7 × 10−10

1.0 × 106

2.7 × 10−5

1.0 × 104

2.7 × 10−7

1.0 × 107

2.7 × 10−4

*

*

*

*

3.0 × 101

8.1 × 102

7.0 × 10−1

1.9 × 101

*

*

*

*

1.0

2.7 × 10-11

1.0 × 103

2.7 × 10-8

*

*

*

*

1.0

2.7× 10-11

1.0 ×103

2.7 × 10-8

*

*

*

*

*

*

bParent

*

*

*

nuclides and their progeny included in secular equilibrium are listed as follows:

Sr-90

Y-90

Zr-93

Nb-93m

Zr-97

Nb-97

Ru-106

Rh-106

Ag-108m

Ag-108

Cs-137

Ba-137m

Ce-144

Pr-144

Ba-140

La-140

Bi-212

Tl-208 (0.36), Po-212 (0.64)

Pb-210

Bi-210, Po-210

Pb-212

Bi-212, Tl-208 (0.36), Po-212 (0.64)

Rn-222

Po-218, Pb-214, Bi-214, Po-214

Ra-223

Rn-219, Po-215, Pb-211, Bi-211, Tl-207

Ra-224

Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

Ra-226

Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210

Ra-228

Ac-228

Th-228

Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212(0.64)

Th-229

Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209

Th-nat

Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212
(0.64)

Th-234

Pa-234m

U-230

Th-226, Ra-222, Rn-218, Po-214

U-232

Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)

U-235

Th-231

U-238

Th-234, Pa-234m

U-nat

Th-234, Pa-234m, U-234, Th-230, Ra-226, Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210,
Bi-210, Po-210

Np-237

Pa-233

Am-242m

Am-242

Am-243

Np-239

cIn

*

the case of Th(nat), the parent nuclide is Th-232; in the case of U(nat), the parent nuclide is U-238.

*

*

*

*
Dated August 22, 2022.

For the Nuclear Regulatory Commission.

Brooke P. Clark,
Secretary of the Commission.

[FR Doc. 2022-18520 Filed: 9/9/2022 8:45 am; Publication Date: 9/12/2022]