[Federal Register Volume 87, Number 175 (Monday, September 12, 2022)]
[Proposed Rules]
[Pages 55708-55734]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 2022-18520]


 ========================================================================
 Proposed Rules
                                                 Federal Register
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 This section of the FEDERAL REGISTER contains notices to the public of 
 the proposed issuance of rules and regulations. The purpose of these 
 notices is to give interested persons an opportunity to participate in 
 the rule making prior to the adoption of the final rules.
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  Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / 
Proposed Rules  

[[Page 55708]]



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 November 28, 2022. 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:
     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: [email protected]. For technical questions contact the 
individual or individuals listed in the FOR FURTHER INFORMATION CONTACT 
section of this document.
     Email comments to: [email protected]. If you do 
not receive an automatic email reply confirming receipt, then contact 
us at 301-415-1677.
     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: 
[email protected]; or Bernard White, 301-415-6577, email: 
[email protected]. 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. Obtaining Information and Submitting Comments
    A. Obtaining Information
    B. Submitting Comments
II. Background
III. Discussion
    A. Action the NRC is Proposing To Take
    B. Applicability of the Proposed Action
    C. Discussion of Issues Specific Request for Comment
IV. Specific Request for Comment
V. Section-by-Section Analysis
VI. Regulatory Flexibility Certification
VII. Regulatory Analysis
VIII. Backfitting and Issue Finality
IX. Cumulative Effects of Regulation
X. Plain Writing
XI. 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:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0179.
     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 [email protected]. For the convenience of the reader, 
instructions about obtaining materials referenced in this document are 
provided in the ``Availability of Documents'' section.
     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 
[email protected] 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

[[Page 55709]]

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 Sec.  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,

[[Page 55710]]

paragraph 417(f). The NRC has determined that the current fissile 
exemptions under Sec.  71.15 provide flexibility for shipping low 
masses or concentrations of fissile materials, and licensees can submit 
a specific exemption request under Sec.  71.12 for fissile materials 
that do not meet the fissile exemption criteria in Sec.  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 dual-purpose 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 Sec.  71.97, stating that there is no actual provision 
requiring advance notification for spent fuel shipments. The 
requirements in Sec.  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. Discussion

A. 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):

[[Page 55711]]



----------------------------------------------------------------------------------------------------------------
        Issue           IAEA  harmonization     DOT  harmonization       Other  changes          No  action
----------------------------------------------------------------------------------------------------------------
                1                      X
                2                                                                                           X
                3                                                                                           X
              4.1                      X
              4.2                      X
                5                      X
                6                      X                      X
                7                      X                      X
                8                      X
                9                      X
               10                      X                      X
               11                      X                      X
               12                                                                    X
               13                                                                    X
               14                                                                    X
             15.1                                                                    X
             15.2                                                                    X
             15.3                      X                      X
             15.4                                                                    X
             15.5                                                                    X
----------------------------------------------------------------------------------------------------------------

Issue 1. Revision of Fissile Exemptions
    The fissile material exemptions in Sec.  71.15 and the fissile 
material general licenses in Sec. Sec.  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 
Sec. Sec.  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 Sec.  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 Sec.  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 (\239\Pu), 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 Sec.  71.15. This limit is based on one 
fifth of a minimum critical mass of uranium-235 (\235\U) (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 \235\U for this limit 
rather than \239\Pu, as any amount of \239\Pu 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 140-gram limit, as with other fissile exemption 
provisions in Sec.  71.15, only relieves the consignor from having to 
ship in a ``Fissile'' package, evaluated per the requirements of 
Sec. Sec.  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 Sec.  71.15(d) for clarity and to 
maintain consistent language throughout Sec.  71.15.
Issue 2. Revision of Reduced External Pressure Test for Normal 
Conditions of Transport
    The regulation at Sec.  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

[[Page 55712]]

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 Sec.  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 Sec.  
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 Sec.  71.55(f), ``General requirements for fissile 
material packages'' and Sec.  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 Sec.  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) Sec. Sec.  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 Sec.  71.71(c)(1), and to add insolation to 
the initial conditions for the tests for hypothetical accident 
conditions in Sec.  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/cm\2\). 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/cm\2\) to metric units (watts per square meter (W/m\2\)). 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 Sec.  71.71(c)(1). Consequently, 
the current units for insolation in 10 CFR part 71 are ``g cal/cm\2\.'' 
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 Sec.  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.

[[Page 55713]]

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 [deg]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 [deg]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 Sec.  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 Sec.  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 Sec.  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 SSR-6. 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 Sec.  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 Sec.  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 
Sec.  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 Sec.  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 Sec.  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 Sec. Sec.  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 
10-3A2/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 Sec.  71.77 and 
make conforming changes to Sec. Sec.  71.4 and 71.100, which both 
reference Sec.  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

[[Page 55714]]

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 Sec.  71.4) could be addressed through a special package 
authorization under Sec.  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 
Sec.  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 Sec.  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 
Sec.  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, Sec.  71.55(g), to 
provide a specific exception for certain uranium hexafluoride 
(UF6) packages from the requirements of Sec.  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 \235\U. To 
use this exception, the applicant must demonstrate, among other things, 
that, following the tests for hypothetical accident conditions in Sec.  
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 x 10-7 ref[middot]cm\3\/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 Sec.  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 Sec.  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 Sec.  71.43(d). The NRC regulations at Sec.  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 Sec.  71.87(b) require that, prior to each shipment, the 
licensee ensures that the package is in unimpaired physical condition 
except

[[Page 55715]]

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 
Sec.  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 non-dispersible 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 Sec.  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 Sec.  71.19 for 
previously-approved packages:
     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;
     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;
     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
     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 Sec.  
71.4 because NRC

[[Page 55716]]

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 Sec.  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 Sec.  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 SSR-6, 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 Sec.  
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 Sec.  71.19(c).
    3. Coordinate with the DOT and make appropriate changes to Sec.  
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 Sec.  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 Sec.  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 Sec.  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 Sec.  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 Sec.  
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 Sec.  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 Sec.  
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

[[Page 55717]]

application. This ensures that all package users, whether NRC licensees 
or not, comply with the requirements listed in Sec.  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 Sec.  
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 Sec.  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 Sec.  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, Sec.  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 Sec. Sec.  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 Sec.  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 Sec.  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 Sec.  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 Sec.  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 Sec.  71.22(d). The 
regulation in Sec.  71.22(e)(1) provides an equation to calculate 
package CSI:
[GRAPHIC] [TIFF OMITTED] TP12SE22.000

where X, Y, and Z are mass limits of \235\U, \233\U, and plutonium 
obtained from Table 71-1 (if \233\U or plutonium are present) or 
Table 71-2.

    Similarly, the general license criteria in Sec.  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 Sec.  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 \239\Pu and \241\Pu 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 Sec.  71.23(d). The regulation in 
Sec.  71.23(e)(1) provides an equation to calculate package CSI:
[GRAPHIC] [TIFF OMITTED] TP12SE22.001

    The calculations that support the mass limits in Sec.  71.22 
include conservative assumptions regarding neutron moderation and water 
reflection, i.e., optimally moderated spheres of \235\U, \233\U, and 
\239\Pu with full water reflection. The mass limits in Sec.  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

[[Page 55718]]

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 \239\Pu in 
Table 71-1 are 37 grams or 24 grams, depending on the degree of 
moderation, while the A2 value for \239\Pu is equivalent to 
0.435 grams). In addition, the requirement in Sec.  71.23 does not 
consistently refer to ``special form sealed sources'' in that paragraph 
(a) also refers to Pu-Be 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 Sec. Sec.  71.22 and 71.23 in a Type B 
package. When shipping material that meets the mass limits of the 
general licenses in Sec. Sec.  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 Sec.  71.0(d)(1). The 
regulations in Sec.  71.0(d)(1) state that use of the general licenses 
in Sec.  71.22 or Sec.  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 Sec.  71.14(b)(1) exempt the licensee from all 
requirements in 10 CFR part 71, except for Sec. Sec.  71.5 and 71.88, 
when shipping a Type A quantity. Because the NRC is proposing to revise 
Sec. Sec.  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 Sec. Sec.  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 Sec.  71.22 or Sec.  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 Sec. Sec.  71.22 and 71.23, the NRC is proposing to 
include three new paragraphs in Sec. Sec.  71.22 and 71.23 that are 
similar to the requirements in Sec.  71.17(c), (d), and (e). The NRC is 
proposing to add a new requirement in Sec. Sec.  71.22(f) and 71.23(f) 
to ensure that, for shipments made using the respective general 
license, each licensee must comply with Sec.  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 
Sec.  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 Sec. Sec.  
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 Sec.  71.17 
or the general license in Sec.  71.22 or Sec.  71.23. Authorizing use 
under the general license in Sec.  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 Sec.  71.22 
or Sec.  71.23.
    Finally, the NRC is proposing to add a new requirement in 
Sec. Sec.  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 Sec.  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 
Sec. Sec.  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 Sec. Sec.  71.22 and 
71.23; and to make conforming changes to Sec.  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 Sec.  71.23.
Issue 14. Deletion of \233\ U Restriction in Fissile General License
    The general license criteria in Sec.  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 
Sec.  71.22(e)(1), and the fissile mass limits in either Table 71-1 or 
71-2 to 10 CFR part 71. Table 71-2 contains mass limits for shipping 
uranium enriched to various weight percent levels in \235\U. However, 
Sec.  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 \235\ U, or if 
\233\ U is present in the package.
    While \233\ U is not present in natural uranium, it may be present 
in very low concentrations in some facilities that may have handled 
\233\ U in the past. These contamination-level concentrations, while 
detectable with modern isotopic assay methods and physically 
``present,'' are not important for criticality safety of \235\ U 
transportation. The calculations used to support the enrichment limit 
for Sec.  71.15(d), for up to 1.0 weight percent enriched uranium, 
demonstrate that this limit is safe provided the plutonium and \233\ U 
are limited to less than one percent of the mass of \235\ U. The same 
limitation could be applied to the use of Table 71-2 limits for 
shipping enriched uranium under Sec.  71.22, without affecting 
criticality safety.
    The NRC is therefore proposing to revise Sec.  71.22 to limit the 
\233\ U to less than one percent of the mass of \235\ U, similar to the 
provision limiting plutonium in Sec.  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

[[Page 55719]]

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 Sec.  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 Sec.  71.4 to change 
``excepted under Sec.  71.15'' to ``exempted under Sec.  71.15.'' This 
change would make the definition of LSA in Sec.  71.4 consistent with 
the title of Sec.  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 Sec.  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-\3\A2 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 x 
10\6\ TBq/g, 1.8 x 10\8\ Ci/g. However, the correct value for the 
specific activity of Rb(nat) is 670 Bq/g (6.7 x 10-\10\ TBq/
g, 1.8 x 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 Sec.  71.21, ``General 
license: Use of foreign approved package''; Sec.  71.22, ``General 
license: Fissile material''; or Sec.  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 Sec.  71.95.
    In the 2004 final rule (69 FR 3697; January 26, 2004) that revised 
Sec.  71.101, ``Quality assurance requirements,'' the NRC stated that 
Sec.  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., 
Sec. Sec.  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

[[Page 55720]]

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 Sec.  71.21, Sec.  
71.22, or Sec.  71.23, which also requires an approved QAP, would be 
impacted.
    Additionally, the regulations in Sec.  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 Sec.  71.95, the NRC stated that 
the compatibility category for Sec.  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 Sec.  
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 
Sec.  71.95 and, therefore, are not required to submit a report to the 
NRC pursuant to Sec.  71.95. The NRC is proposing to change the 
compatibility category for Sec.  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 Sec.  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 Sec.  71.95(b) so that Agreement 
States receive these QAP-related reports. The compatibility categories 
for Sec.  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) Sec.  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., 
Sec. Sec.  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 Sec.  
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 Sec.  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 Sec.  71.95(b), and to Sec.  
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 Sec.  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 Sec.  
73.72. Additionally, as required in Sec.  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 Sec.  
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 Sec.  71.97. Therefore, licensees may need to produce two 
reports for a single shipment to meet the advance notification 
requirements of Sec. Sec.  71.97 and 73.37 or Sec.  73.35. To address 
this potential inefficiency the NRC is proposing to modify Sec.  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

[[Page 55721]]

Exceptions in IAEA SSR-6, paragraph 417; Issue No. 1b: Competent 
Authority-Approved 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 Sec. Sec.  71.55 
and 71.59 is deemed too burdensome, the licensee could request a 
specific exemption under Sec.  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 NRC-issued 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 Sec. Sec.  71.22 and 71.23 are appropriate for 
providing criticality safety equivalent to packages approved under the 
criticality safety requirements of Sec. Sec.  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 Sec.  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 Sec.  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 (Sec.  71.15) and fissile material 
general licenses (Sec. Sec.  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 Sec.  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 Sec.  
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 (Sec. Sec.  71.17(c) and 71.87(b)) ensure that, prior 
to transport, the licensee must comply with the terms and conditions

[[Page 55722]]

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 NRC-approved 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.
     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.
     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 Sec.  71.106 to achieve NRC's stated intent in 
the 2015 final rule. Specifically, the NRC is proposing to revise Sec.  
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 as-needed 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.
     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.

[[Page 55723]]

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 ``Sec. Sec.  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 Sec. Sec.  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 \233\U to less than one 
percent of the mass of \235\U, similar to the provision limiting 
plutonium in Sec.  71.22(e)(5)(ii), and add new paragraphs (f) through 
(h) to ensure that each licensee will comply with Sec.  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 Sec.  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 Sec.  71.23.

Section 71.31 Contents of Application

    This proposed rule would revise paragraph (a) to add a maintenance 
program description, as required by Sec.  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 Sec.  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 Sec. Sec.  
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/cm\2\)'' to ``(W/m\2\)''.

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 Sec.  71.77 and make 
conforming changes to Sec. Sec.  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 Sec.  71.97 with those in 
Sec. Sec.  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 Sec.  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 (Sec.  2.810). However, as indicated in the regulatory analysis, 
these amendments do not have a significant economic impact on the 
affected small entities.

[[Page 55724]]

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 (Sec.  50.109, Sec.  70.76, 
Sec.  72.62, or Sec.  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

[[Page 55725]]

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: 1,376.7 hours (an increase of 
1,052.5 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:
     Federal Rulemaking Website: Go to https://www.regulations.gov and search for Docket ID NRC-2016-0179.
     Mail comments to: FOIA, Library, and Information 
Collections Branch T6-A10M, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, or by email to 
[email protected].
     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 November 14, 2022. 
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 Sec.  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 Sec.  71.77 and 
replace the section heading with ``RESERVED.'' This change would impact 
Sec.  71.100(b), because Sec.  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

[[Page 55726]]

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., Sec. Sec.  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 Sec.  
71.21, Sec.  71.22 or Sec.  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 Sec.  34.31), or that ship using the general license in 
Sec.  71.21, Sec.  71.22, or Sec.  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.

----------------------------------------------------------------------------------------------------------------
                                                                                          Compatibility
             Section                     Change                Subject         ---------------------------------
                                                                                    Existing           New
----------------------------------------------------------------------------------------------------------------
71.0(d)(1)......................  Revised............  Purpose and Scope......  D                D
71.4............................  New................  Definition: Radiation    ...............  [A]
                                                        Level.
71.4............................  Revised............  Definition: Low          [B]              [B]
                                                        Specific Activity
                                                        (LSA) material
                                                        [Deletion of Low
                                                        Specific Activity--III
                                                        Leaching Test].
71.4............................  Revised............  Definition: Special      [B]              [B]
                                                        form radioactive
                                                        material.
71.4............................  Revised............  Definition: Surface      [B]              [B]
                                                        Contaminated Object
                                                        (SCO).
71.15(a) and (d)................  Revised............  Exemption from           [B]              [B]
                                                        classification as
                                                        fissile material.
71.15(g)........................  New................  Exemption from           ...............  [B]
                                                        classification as
                                                        fissile material.
71.17(e)........................  Revised............  General license: NRC-    B                B
                                                        approved package.
71.19...........................  Revised............  Previously approved      NRC              NRC
                                                        package.
71.22(a), (c), and (e)(3)         Revised............  General license:         [B]              [B]
 through (5).                                           Fissile material.
71.22(f) through (h)............  New................  General license:         ...............  [B]
                                                        Fissile material.
71.23(a) and (c)................  Revised............  General license:         [B]              [B]
                                                        Plutonium-beryllium
                                                        special form material.
71.23(f) through (h)............  New................  General license:         ...............  [B]
                                                        Plutonium-beryllium
                                                        special form material.
71.31(a)........................  Revised............  Contents of application  NRC              NRC
71.35(b) and (c)................  Revised............  Package evaluation.....  NRC              NRC
71.35(d)........................  New................  Package evaluation.....  ...............  NRC
71.43(d)........................  Revised............  General standards for    NRC              NRC
                                                        all packages.
71.43(i)........................  New................  General standards for    ...............  NRC
                                                        all packages.
71.55(g)........................  Revised............  General requirements     NRC              NRC
                                                        for fissile material
                                                        packages.
71.71(c)(1).....................  Revised............  Normal conditions of     NRC              NRC
                                                        transport.
71.73(b)........................  Revised............  Hypothetical accident    NRC              NRC
                                                        conditions.
71.77...........................  Removed............  Qualification of LSA--   NRC              ...............
                                                        III Material.
71.95...........................  Revised              Reports................  D                ** C
                                   compatibility
                                   category.
71.95(a)(3).....................  Removed............  Reports................  D                *
71.97...........................  Revised............  Advance notification of  B                B
                                                        shipment of irradiated
                                                        reactor fuel and
                                                        nuclear waste.
71.100..........................  Revised............  Criminal penalties.....  D                D
71.101(b).......................  Revised              Quality assurance        *** C            *** B
                                   compatibility        requirements.
                                   category.
71.101(c)(1)....................  Revised              Quality assurance        *** C            ** B
                                   compatibility        requirements.
                                   category.
71.103(a) and (b)...............  Revised              Quality assurance        *** C            ** B
                                   compatibility        organization.
                                   category.
71.103(c), (d), (e) and (f).....  Revised              Quality assurance        D                ** B
                                   compatibility        organization.
                                   category.
71.105..........................  Revised              Quality assurance        C                ** B
                                   compatibility        program.
                                   category.
71.106..........................  Revised              Changes to quality       C                ** B
                                   compatibility        assurance program.
                                   category.
71.109..........................  Revised              Procurement document     NRC              ** B
                                   compatibility        control.
                                   category.

[[Page 55727]]

 
71.111..........................  Revised              Instructions,            NRC              ** B
                                   compatibility        procedures and
                                   category.            drawings.
71.113..........................  Revised              Document control.......  NRC              ** B
                                   compatibility
                                   category.
71.115..........................  Revised              Control of purchased     NRC              ** B
                                   compatibility        material, equipment,
                                   category.            and services.
71.117..........................  Revised              Identification and       NRC              ** B
                                   compatibility        control of materials,
                                   category.            parts and components.
71.119..........................  Revised              Control of special       NRC              ** B
                                   compatibility        processes.
                                   category.
71.121..........................  Revised              Internal inspection....  NRC              ** B
                                   compatibility
                                   category.
71.123..........................  Revised              Test control...........  NRC              ** B
                                   compatibility
                                   category.
71.125..........................  Revised              Control of measuring     NRC              ** B
                                   compatibility        and test equipment.
                                   category.
71.127..........................  Revised              Handling, storage, and   [C]              ** B
                                   compatibility        shipping control.
                                   category.
71.129..........................  Revised              Inspection, test, and    [C]              ** B
                                   compatibility        operating status.
                                   category.
71.131..........................  Revised              Nonconforming            [C]              ** B
                                   compatibility        materials, parts, or
                                   category.            components.
71.133..........................  Revised              Corrective action......  C                ** B
                                   compatibility
                                   category.
71.135..........................  Revised              Quality assurance        *** C            ** C
                                   compatibility        records.
                                   category.
71.137..........................  Revised              Audits.................  C                ** C
                                   compatibility
                                   category.
Table A-1 in Appendix A to 10     Revised............  A1 and A2 Values for     [B]              [B]
 CFR Part 71.                                           Radionuclides.
Table A-2 in Appendix A to 10     Revised............  Exempt Material          [B]              [B]
 CFR Part 71.                                           Activity
                                                        Concentrations and
                                                        Exempt Consignment
                                                        Activity Limits for
                                                        Radionuclides.
----------------------------------------------------------------------------------------------------------------
* 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 Sec.   71.21,
  Sec.   71.22, or Sec.   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
  Sec.   34.31(b). It also indicated that this section satisfies Sec.   71.17(b) and therefore will satisfy
  those sections referenced in this provision (Sec.  Sec.   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, Public Law 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

[[Page 55728]]

searching on https://www.regulations.gov under Docket ID NRC-2016-0179. 
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 website at least 10 calendar days before the meeting. 
Stakeholders should monitor the NRC's public meeting website 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.

------------------------------------------------------------------------
                                                ADAMS accession No./web
                   Document                      link/ Federal Register
                                                        citation
------------------------------------------------------------------------
                   Rulemaking Documents and References
------------------------------------------------------------------------
SECY-20-0102 for this proposed rule..........  ML20101F921
Federal Register notice for this proposed      ML22209A035
 rule.
Regulatory Analysis for this proposed rule...  ML22209A039
Environmental Assessment for this proposed     ML22209A045
 rule.
OMB supporting statement for this proposed     ML22209A052
 rule.
Draft regulatory basis document for this       ML18262A185
 rulemaking, dated March 2019.
Federal Register notification for draft        84 FR 14898
 regulatory basis, dated April 12, 2019.
Draft regulatory basis comment submission #1.  ML19106A347
Draft regulatory basis comment submission #2.  ML19113A064
Draft regulatory basis comment submission #3.  ML19143A311
Draft regulatory basis comment submission #4.  ML19143A312
Draft regulatory basis comment submission #5.  ML19148A147
Draft regulatory basis comment submission #6.  ML19149A474
Draft regulatory basis comment submission #7.  ML19150A140
NRC final rule amending packaging and          80 FR 33988
 transportation of radioactive material
 regulations, dated June 12, 2015.
DOT final rule amending packaging and          79 FR 40589
 transportation of radioactive material
 regulations, dated July 11, 2014.
NRC final rule harmonizing its regulations     69 FR 3697
 with the 1996 edition of IAEA Safety Series
 No. 6, dated January 26, 2004.
NRC proposed rule harmonizing its regulations  67 FR 21390
 with the 1996 edition of IAEA Safety Series
 No. 6, dated April 30, 2002.
NRC final rule harmonizing its regulations     60 FR 50248
 with the 1985 edition of IAEA Safety Series
 No. 6, dated September 28, 1995.
NRC/DOT Memorandum of Understanding, dated     44 FR 38690
 July 2, 1979.
SECY-16-0093, ``Rulemaking Plan for Revisions  ML16158A164
 to Transportation Safety Requirements and
 Harmonization with International Atomic
 Energy Agency Transportation Requirements,''
 dated July 28, 2016.

[[Page 55729]]

 
Staff Requirements Memorandum SRM-SECY-16-     ML16235A182
 0093, ``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  ML16299A298 paper,
 Potential Revisions to Transportation Safety   ML16299A291 package
 Requirements and Harmonization with
 International Atomic Energy Agency
 Transportation Requirements,'' dated
 November 15, 2016.
Federal Register notification for              81 FR 83171
 harmonization issues paper, dated November
 21, 2016.
Issues paper public meeting summary,           ML16343A661
 ``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     ML22223A085
 Format and Content of Part 71 Applications
 for Approval of Packages for Radioactive
 Material,'' Revision 3 of Regulatory Guide
 7.9.
------------------------------------------------------------------------
       IAEA Transportation Safety Standards and Related References
------------------------------------------------------------------------
SSR-6, ``Regulations for the Safe Transport    https://www.iaea.org/
 of Radioactive Material,'' 2018 Edition.       publications/12288/
                                                regulations-for-the-safe-
                                                transport-of-radioactive-
                                                material
SSR-6, ``Regulations for the Safe Transport    https://www.iaea.org/
 of Radioactive Material,'' 2012 Edition.       publications/8851/
                                                regulations-for-the-safe-
                                                transport-of-radioactive-
                                                material-2012-edition
TS-R-1, ``Regulations for the Safe Transport   https://www.iaea.org/
 of Radioactive Material,'' 2009 Edition.       publications/8005/
                                                regulations-for-the-safe-
                                                transport-of-radioactive-
                                                material-2009-edition
TS-R-1, ``Regulations for the Safe Transport   https://www.iaea.org/
 of Radioactive Material,'' 2005 Edition.       publications/7291/
                                                regulations-for-the-safe-
                                                transport-of-radioactive-
                                                material-2005-edition
TS-R-1, ``Regulations for the Safe Transport   https://www.iaea.org/
 of Radioactive Material,'' 1996 Edition.       publications/6056/
                                                regulations-for-the-safe-
                                                transport-of-radioactive-
                                                material-1996-edition-
                                                revised
Safety Series No. 6, ``Regulations for the     http://gnssn.iaea.org/
 Safe Transport of Radioactive Material, 1985   Superseded%20Safety%20St
 Edition (As Amended in 1990)''.                andards/
                                                Safety_Series_006_1990.p
                                                df
Safety Series No. 6, ``Regulations for the     https://gnssn.iaea.org/
 Safe Transport of Radioactive Material,''      Superseded%20Safety%20St
 1985 Edition.                                  andards/
                                                Safety_Series_006_1985.p
                                                df
Safety Series No. 6, ``Regulations for the     https://gnssn.iaea.org/
 Safe Transport of Radioactive Material,''      Superseded%20Safety%20St
 1973 Edition.                                  andards/
                                                Safety_Series_006_1973.p
                                                df
Safety Series No. 6, ``Regulations for the     https://gnssn.iaea.org/
 Safe Transport of Radioactive Material,''      Superseded%20Safety%20St
 1967 Edition.                                  andards/
                                                Safety_Series_006_1967.p
                                                df
------------------------------------------------------------------------
                Other International Standards References
------------------------------------------------------------------------
ANSI N14.1-2012, ``Nuclear Materials--Uranium  https://webstore.ansi.org/
 Hexafluoride--Packagings for Transport,''      standards/pcc/
 dated December 3, 2012.                        ansin142012
ANSI N14.5-2014, ``American National Standard  https://webstore.ansi.org/
 for Radioactive Materials--Leakage Tests on    standards/pcc/
 Packages for Shipment,'' dated June 19, 2014.  ansin142014
International Organization for                 https://www.iso.org/
 Standardization 7195:2005, ``Nuclear Energy--  standard/31251.html
 Packaging of Uranium Hexafluoride (UF6) for
 Transport,'' dated September 2005.
American National Standards Institute/         https://webstore.ansi.org/
 American Nuclear Society 8.1-2014              Standards/ANSI/
 (Reaffirmed 2018), ``Nuclear Criticality       ANSIANS2014R2018
 Safety in Operations with Fissionable
 Materials Outside Reactors,'' American
 Nuclear Society, La Grange Park, IL.
------------------------------------------------------------------------
                        Miscellaneous References
------------------------------------------------------------------------
National Renewable Energy Laboratory Solar     https://www.nrel.gov/gis/
 Radiation Data.                                assets/images/solar-
                                                annual-ghi-2018-usa-
                                                scale-01.jpg
NRC letter to Agreement States,                ML17213A844
 ``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   63 FR 31885
 Government Writing,'' published June 10,
 1998.
Agreement State Program Policy Statement,      82 FR 48535
 dated October 18, 2017.
NRC Management Directive 5.9, Handbook 5.9,    ML18081A070
 ``Adequacy and Compatibility of Program
 Elements for Agreement State Programs,''
 dated April 26, 2018.
NRC Management Directive 8.4, ``Management of  ML18093B087
 Backfitting, Forward Fitting, Issue
 Finality, and Information Requests,'' dated
 September 20, 2019.

[[Page 55730]]

 
ORNL/TM-2014/658, ``Comparison of the          https://rampac.energy.gov/
 International and United States Domestic       docs/default-source/
 Radioactive Material Transport                 doeinfo/ORNL-TM-2014-
 Regulations,'' dated September 30, 2014.       658.pdf
NUREG-1409, ``Backfitting Guidelines,''        ML18109A498
 Revision 1, draft for public comment, dated
 March 2020.
NUREG-1608, ``Categorizing and Transporting    ML15336A927
 Low Specific Activity Materials and Surface
 Contaminated Objects,'' dated July 1998.
NUREG-1660, ``U.S.-Specific Schedules of       https://rampac.energy.gov/
 Requirements for Transport of Specified        docs/default-source/
 Types of Radioactive Material                  nrcinfo/nureg_1660.pdf
 Consignments,'' dated January 1999.
NUREG-1886, ``Joint Canada-United States       ML090930197
 Guide for Approval of Type B(U) and Fissile
 Material Transportation Packages,'' dated
 March 2009.
NUREG-2216, ``Standard Review Plan for         ML20234A651
 Transportation Packages for Spent Fuel and
 Radioactive Material,'' dated August 2020.
------------------------------------------------------------------------

    Throughout the development of this proposed rule, the NRC may post 
documents related to it, including public comments, on the Federal 
rulemaking website 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

0
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).

0
2. In Sec.  71.0, revise paragraph (d)(1) to read as follows:


Sec.  71.0   Purpose and scope.

* * * * *
    (d)(1) Exemptions from the requirement for license in Sec.  71.3 
are specified in Sec.  71.14. The general license in Sec.  71.21 does 
not require NRC package approval. The general licenses in Sec. Sec.  
71.22 and 71.23 require NRC package approval if the quantities exceed a 
Type A quantity. The general license in Sec.  71.17 requires that an 
NRC certificate of compliance or other package approval be issued for 
the package to be used under this general license.
* * * * *
0
3. Amend Sec.  71.4 by:
0
a. Revising the definitions for Low Specific Activity material and 
Special form radioactive material;
0
b. Revising the introductory text and add paragraph (3) for Surface 
contaminated object; and
0
c. Adding the definition Radiation level in alphabetical order.
    The revisions and addition read as follows:


Sec.  71.4   Definitions.

* * * * *
    Low Specific Activity (LSA) material means radioactive material 
with limited specific activity which is nonfissile or is exempt under 
Sec.  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 Sec.  71.75. A special form 
encapsulation designed in accordance with the requirements of Sec.  
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 Sec.  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

[[Page 55731]]

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 cm\2\ does not exceed 8 x 10\5\ 
Bq/cm\2\ (20 microcuries/cm\2\) for beta and gamma emitters and low 
toxicity alpha emitters, or 8 x 10\4\ Bq/cm\2\ (2 microcuries/cm\2\) 
for all other alpha emitters.
* * * * *
0
4. In Sec.  71.15, revise the introductory text and paragraphs (a) and 
(d) and add paragraph (g) to read as follows:


Sec.  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 Sec. Sec.  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 uranium-235 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.
0
5. In Sec.  71.17, revise paragraph (e) to read as follows:


Sec.  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 Sec.  71.19.
0
6. Amend Sec.  71.19 by:
0
a. Revising paragraph (a);
0
b. Redesignating paragraphs (c) and (d) as paragraphs (d) and (e);
0
c. Adding new paragraph (c); and
0
d. Revising newly redesignated paragraph (e).
    The revisions and addition read as follows:


Sec.  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 Sec.  71.17 with the following 
additional conditions:
    (1) Fabrication of the package is satisfactorily completed by April 
1, 1999, as demonstrated by application of its model number in 
accordance with Sec.  71.85(c);
    (2) A 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 Sec.  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 Sec.  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 AF-96 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.
0
7. In Sec.  71.22, revise paragraphs (a), (c), and (e)(3) through (5) 
and add paragraphs (f) through (h) to read as follows:


Sec.  71.22   General license: Fissile material.

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material, or to deliver fissile material to a 
carrier for transport, if the material is shipped in accordance with 
this section. The fissile material need not be contained in a package 
which meets the standards of Sec. Sec.  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 
Sec.  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

[[Page 55732]]

which a license, CoC, or other approval has been issued by the NRC, and 
must comply with the provisions in Sec.  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 Sec.  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 Sec.  71.19.
0
8. In Sec.  71.23, revise paragraph (a) and the introductory text of 
paragraph (c) and add paragraphs (f) through (h) to read as follows:


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

    (a) A general license is issued to any licensee of the Commission 
to transport fissile material in the form of plutonium-beryllium (Pu-
Be) special form 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 Sec. Sec.  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 Sec.  
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 Sec.  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 Sec.  71.19.
0
9. In Sec.  71.31, revise paragraph (a) to read as follows:


Sec.  71.31   Contents of application.

    (a) An application for an approval under this part must include, 
for each proposed packaging design, the following information:
    (1) A package description as required by Sec.  71.33;
    (2) A package evaluation as required by Sec.  71.35;
    (3) A maintenance program description, as required by Sec.  71.35; 
and
    (4) A quality assurance program description, as required by Sec.  
71.37, or a reference to a previously approved quality assurance 
program.
* * * * *
0
10. In Sec.  71.35, revise paragraphs (b) and (c) and add paragraph (d) 
to read as follows:


Sec.  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 
Sec.  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.
0
11. In Sec.  71.43, revise paragraph (d) and add paragraph (i) to read 
as follows:


Sec.  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 Sec.  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 Sec. Sec.  71.71 and 71.73, there would be adequate space 
to accommodate variations in temperature of the liquid, dynamic 
effects, and filling dynamics.
0
12. In Sec.  71.55, revise paragraph (g)(1) to read as follows:


Sec.  71.55   General requirements for fissile material packages.

* * * * *
    (g) * * *
    (1) Following the tests specified in Sec.  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;
* * * * *
0
13. In Sec.  71.71, in the table in paragraph (c)(1), revise the 
heading of the second column to read as follows:


Sec.  71.71   Normal conditions of transport.

* * * * *
    (c) * * *
    (1) * * *

                             Insolation Data
------------------------------------------------------------------------
 
------------------------------------------------------------------------
* * *.....................................  Total insolation for a 12-
                                             hour period (W/m\2\)
 
                                * * * * *
------------------------------------------------------------------------

* * * * *
0
14. In Sec.  71.73, revise paragraph (b) to read as follows:


Sec.  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 [deg]C (-20 [deg]F) and +38 [deg]C (+100 [deg]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 Sec.  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.
* * * * *


Sec.  71.77   [Removed and Reserved]

0
15. Remove and reserve Sec.  71.77.

[[Page 55733]]

Sec.  71.95   [Amended]

0
16. In Sec.  71.95, remove paragraph (a)(3).


Sec.  71.97   [Amended]

0
17. In Sec.  71.97:
0
a. In the section heading, remove the phrase ``irradiated reactor fuel 
and'';
0
b. In paragraph (b) introductory text, remove the word ``also'';
0
c. In paragraph (d) introductory text and paragraphs (d)(1) and (2), 
remove the phrase ``irradiated reactor fuel or''; and
0
d. In paragraph (f)(1), remove the phrase ``an irradiated reactor fuel 
or'' and add in its place the word ``a''.


Sec.  71.100   [Amended]

0
18. In Sec.  71.100(b), remove the reference ``71.77,''.
0
19. In Sec.  71.106, revise the introductory text of paragraph (b) to 
read as follows:


Sec.  71.106   Changes to quality assurance program.

* * * * *
    (b) Each quality assurance program approval holder may change a 
previously approved quality assurance program without prior NRC 
approval, if the change does not reduce the commitments in the quality 
assurance program previously approved by the NRC. Changes to the 
quality assurance program that do not reduce the commitments shall be 
submitted to the NRC every 24 months, in accordance with Sec.  71.1(a). 
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 Sec.  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:
* * * * *
0
20. In appendix A to part 71, in paragraph V.b.:
0
a. In Table A-1, add the entries for Ba-135m, Ge-69, Ir-193m, Ni-57, 
Sr-83, Tb-149, and Tb-161 in alphanumeric order and revise the entries 
for Ni-59, Rb(nat), and Tb-157; and
0
b. In Table A-2, add the entries for Ba-135m, Ge-69, Ir-193m, Ni-57, 
Sr-83, Tb-149, 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
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                                                                                Specific activity
   Symbol of  radionuclide        Element and        A1 (TBq)         A1 (Ci)\b\         A2 (TBq)         A2 (Ci)\b\   ---------------------------------
                                 atomic number                                                                              (TBq/g)           (Ci/g)
--------------------------------------------------------------------------------------------------------------------------------------------------------
 
                                                                      * * * * * * *
Ba-135m......................  ................  2.0 x 10\1\.....  5.4 x 10\2\.....  6.0 x 10-\1\....  1.6 x 10\1\....  3.0 x 10\4\....  8.1 x 10\5\
 
                                                                      * * * * * * *
Ge-69........................  ................  1.0 x 10\0\.....  2.7 x 10\1\.....  1.0 x 10\0\.....  2.7 x 10\1\....  4.3 x 10\4\....  1.2 x 10\6\
 
                                                                      * * * * * * *
Ir-193m......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\0\.....  1.1 x 10\2\....  2.4 x 10\3\....  6.4 x 10\4\
 
                                                                      * * * * * * *
Ni-57........................  Nickel (28).....  6.0 x 10-\1\....  1.6 x 10\1\.....  5.0 x 10-\1\....  1.4 x 10\1\....  5.7 x 10\4\....  1.5 x 10\6\
Ni-59........................  ................  Unlimited.......  Unlimited.......  Unlimited.......  Unlimited......  3.0 x 10-\3\...  8.0 x 10-\2\
 
                                                                      * * * * * * *
Rb(nat)......................  ................  Unlimited.......  Unlimited.......  Unlimited.......  Unlimited......  6.7 x 10-\10\..  1.8 x 10-\8\
 
                                                                      * * * * * * *
Sr-83........................  ................  1.0 x 10\0\.....  2.7 x 10\1\.....  1.0 x 10\0\.....  2.7 x 10\1\....  4.3 x 10\4\....  1.2 x 10\6\
 
                                                                      * * * * * * *
Tb-149.......................  Terbium (65)....  8.0 x 10-\1\....  2.2 x 10\1\.....  8.0 x 10-\1\....  2.2 x 10\1\....  1.9 x 10\5\....  5.1 x 10\6\
Tb-157.......................  ................  4.0 x 10\1\.....  1.1 x 10\3\.....  4.0 x 10\1\.....  1.1 x 10\3\....  5.6 x 10-\1\...  1.5 x 10\1\
 
                                                                      * * * * * * *
Tb-161.......................  ................  3.0 x 10\1\.....  8.1 x 10\2\.....  7.0 x 10-\1\....  1.9 x 10\1\....  4.3 x 10\3\....  1.2 x 10\5\
 
                                                                      * * * * * * *
--------------------------------------------------------------------------------------------------------------------------------------------------------


                       Table A-2--Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides
--------------------------------------------------------------------------------------------------------------------------------------------------------
                                                                    Activity                Activity
                                       Element and  atomic      concentration for      concentration for     Activity limit  for    Activity limit  for
      Symbol of  radionuclide                number           exempt material  (Bq/  exempt material  (Ci/   exempt  consignment    exempt  consignment
                                                                       g)                      g)                    (Bq)                   (Ci)
--------------------------------------------------------------------------------------------------------------------------------------------------------
 
                                                                      * * * * * * *
Ba-135m............................  ......................  1.0 x 10\2\...........  2.7 x 10-\9\.........  1.0 x 10\6\..........  2.7 x 10-\5\
 
                                                                      * * * * * * *
Ge-69..............................  ......................  1.0 x 10\1\...........  2.7 x 10-\10\........  1.0 x 10\6\..........  2.7 x 10-\5\
 
                                                                      * * * * * * *
Ir-193m............................  ......................  1.0 x 10\4\...........  2.7 x 10-\7\.........  1.0 x 10\7\..........  2.7 x 10-\4\
 

[[Page 55734]]

 
                                                                      * * * * * * *
Ni-57..............................  Nickel (28)...........  1.0 x 10\1\...........  2.7 x 10-\10\........  1.0 x 10\6\..........  2.7 x 10-\5\
Ni-59..............................  ......................  1.0 x 10\4\...........  2.7 x 10-\7\.........  1.0 x 10\8\..........  2.7 x 10-\3\
 
                                                                      * * * * * * *
Sr-83..............................  ......................  1.0 x 10\1\...........  2.7 x 10-\10\........  1.0 x 10\6\..........  2.7 x 10-\5\
 
                                                                      * * * * * * *
Tb-149.............................  Terbium (65)..........  1.0 x 10\1\...........  2.7 x 10-\10\........  1.0 x 10\6\..........  2.7 x 10-\5\
Tb-157.............................  ......................  1.0 x 10\4\...........  2.7 x 10-\7\.........  1.0 x 10\7\..........  2.7 x 10-\4\
 
                                                                      * * * * * * *
Tb-161.............................  ......................  3.0 x 10\1\...........  8.1 x 10\2\..........  7.0 x 10-\1\.........  1.9 x 10\1\
 
                                                                      * * * * * * *
Th(nat) (b), (c)...................  ......................  1.0...................  2.7 x 10-\11\........  1.0 x 10\3\..........  2.7 x 10-\8\
 
                                                                      * * * * * * *
U(nat) (b), (c)....................  ......................  1.0...................  2.7 x 10-\11\........  1.0 x 10\3\..........  2.7 x 10-\8\
 
                                                                      * * * * * *
--------------------------------------------------------------------------------------------------------------------------------------------------------
* * * * *
\b\ Parent nuclides and their progeny included in secular equilibrium are listed as follows:


------------------------------------------------------------------------
 
------------------------------------------------------------------------
Sr-90.............................  Y-90
Zr-93.............................  Nb-93m
Zr-97.............................  Nb-97
Ru-106............................  Rh-106
Ag-108m...........................  Ag-108
Cs-137............................  Ba-137m
Ce-144............................  Pr-144
Ba-140............................  La-140
Bi-212............................  Tl-208 (0.36), Po-212 (0.64)
Pb-210............................  Bi-210, Po-210
Pb-212............................  Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-222............................  Po-218, Pb-214, Bi-214, Po-214
Ra-223............................  Rn-219, Po-215, Pb-211, Bi-211, Tl-
                                     207
Ra-224............................  Rn-220, Po-216, Pb-212, Bi-212, Tl-
                                     208 (0.36), Po-212 (0.64)
Ra-226............................  Rn-222, Po-218, Pb-214, Bi-214, Po-
                                     214, Pb-210, Bi-210, Po-210
Ra-228............................  Ac-228
Th-228............................  Ra-224, Rn-220, Po-216, Pb-212, Bi-
                                     212, Tl-208 (0.36), Po-212(0.64)
Th-229............................  Ra-225, Ac-225, Fr-221, At-217, Bi-
                                     213, Po-213, Pb-209
Th-nat............................  Ra-228, Ac-228, Th-228, Ra-224, Rn-
                                     220, Po-216, Pb-212, Bi-212, Tl-208
                                     (0.36), Po-212 (0.64)
Th-234............................  Pa-234m
U-230.............................  Th-226, Ra-222, Rn-218, Po-214
U-232.............................  Th-228, Ra-224, Rn-220, Po-216, Pb-
                                     212, Bi-212, Tl-208 (0.36), Po-212
                                     (0.64)
U-235.............................  Th-231
U-238.............................  Th-234, Pa-234m
U-nat.............................  Th-234, Pa-234m, U-234, Th-230, Ra-
                                     226, Rn-222, Po-218, Pb-214, Bi-
                                     214, Po-214, Pb-210, Bi-210, Po-210
Np-237............................  Pa-233
Am-242m...........................  Am-242
Am-243............................  Np-239
------------------------------------------------------------------------
\c\ In 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 9-9-22; 8:45 am]
BILLING CODE 7590-01-P