Harmonization of Transportation Safety Requirements With IAEA Standards, 55708-55734 [2022-18520]
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55708
Proposed Rules
Federal Register
Vol. 87, No. 175
Monday, September 12, 2022
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.
NUCLEAR REGULATORY
COMMISSION
10 CFR Part 71
[NRC–2016–0179]
RIN 3150–AJ85
Harmonization of Transportation
Safety Requirements With IAEA
Standards
Nuclear Regulatory
Commission.
ACTION: Proposed rule and guidance;
request for comment.
AGENCY:
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.’’
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
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FOR FURTHER INFORMATION CONTACT
section of this document.
• Email comments to:
Rulemaking.Comments@nrc.gov. If you
do not receive an automatic email reply
confirming receipt, then contact us at
301–415–1677.
• 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:
SUMMARY:
DATES:
Forder; telephone: 301–415–3407;
email: Dawn.Forder@nrc.gov. For
technical questions contact the
individual or individuals listed in the
James Firth, 301–415–6628, email:
James.Firth@nrc.gov; or Bernard White,
301–415–6577, email: Bernard.White@
nrc.gov. Both are staff of the Office of
Nuclear Material Safety and Safeguards,
U.S. Nuclear Regulatory Commission,
Washington, DC 20555–0001.
SUPPLEMENTARY INFORMATION:
Table of Contents
I. 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
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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 publiclyavailable 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 publiclyavailable documents online in the
ADAMS Public Documents collection at
https://www.nrc.gov/reading-rm/
adams.html. To begin the search, select
‘‘Begin Web-based ADAMS Search.’’ For
problems with ADAMS, please contact
the NRC’s Public Document Room (PDR)
reference staff at 1–800–397–4209, 301–
415–4737, or by email to
PDR.Resource@nrc.gov. For the
convenience of the reader, instructions
about obtaining materials referenced in
this document are provided in the
‘‘Availability of Documents’’ section.
• NRC’s PDR: You may examine and
purchase copies of public documents,
by appointment, at the PDR, Room P1
B35, One White Flint North, 11555
Rockville Pike, Rockville, Maryland
20852. To make an appointment to visit
the PDR, please send an email to
PDR.Resource@nrc.gov or call 1–800–
397–4209 or 301–415–4737, between
8:00 a.m. and 4:00 p.m. eastern time
(ET), Monday through Friday, except
Federal holidays.
B. Submitting Comments
Please include Docket ID NRC–2016–
0179 in your comment submission.
The NRC cautions you not to include
identifying or contact information that
you do not want to be publicly
disclosed in your comment submission.
The NRC will post all comment
submissions at https://
www.regulations.gov as well as enter the
comment submissions into ADAMS.
The NRC does not routinely edit
comment submissions to remove
identifying or contact information.
If you are requesting or aggregating
comments from other persons for
submission to the NRC, then you should
inform those persons not to include
identifying or contact information that
they do not want to be publicly
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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
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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—
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issue license-specific conditions and
exemptions; and Alternative 4–initiate a
rulemaking action to revise 10 CFR part
71. The alternatives were evaluated
based on their viability to resolve the
regulatory issues of concern and
estimates of their costs and potential
benefits. The NRC determined that the
rulemaking action, Alternative 4, for
Issues 1 (in part), 2, and 4–15, in
combination with the no-action
alternative, Alternative 1, for Issue 3,
was the NRC-recommended action
because it represented the most effective
and least-costly option. Alternatives 2
and 3 would not address all of the
regulatory issues or would result in
higher costs to the NRC and industry.
The NRC also held a public meeting
on April 30, 2019, to discuss the draft
regulatory basis and answer questions.
The NRC received seven public
comment submissions on the draft
regulatory basis—three with general
comments on the rulemaking and four
with comments on specific issues—as
well as comments that were considered
outside the scope of this proposed rule.
All three general comments were
supportive of the harmonization effort
with IAEA SSR–6. The NRC did not
receive any comments on Issues 2, 6,
and 14. The NRC received comments
supportive of the proposal for Issues 4b,
11, 12, 13 and 15, along with comments
supportive of other issues which also
recommended modifications to the
NRC’s proposed changes. One comment
on Issue 5 proposed the NRC add a
definition of ‘‘radiation level’’ to 10 CFR
part 71, which the NRC included in this
proposed rule.
One comment on Issue 1 stated that
the fissile exemption mass limits in 10
CFR part 71 should match those in SSR–
6, paragraph 417, to avoid confusion for
international shipments from the United
States. The NRC has determined that its
regulations for fissile exemption mass
limits should differ from the IAEA’s
requirements to provide flexibility for
shippers. Specifically, the NRC
requirements in this proposed rule
would adopt a 3.5-gram limit from SSR–
6, paragraph 417(c), but without the
associated consignment limit found in
paragraph 570(c); they also would adopt
a higher mass limit than SSR–6,
paragraph 417(e). Several existing fissile
exemptions under § 71.15 do not have
corresponding exceptions under SSR–6,
paragraph 417; if the NRC made 10 CFR
part 71 fissile exemptions identical to
the fissile exceptions in SSR–6,
paragraph 417, fissile material licensees
would lose the benefit of these
exemptions. Also, the NRC is not
pursuing the competent authorityapproved exception in SSR–6,
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paragraph 417(f). The NRC has
determined that the current fissile
exemptions under § 71.15 provide
flexibility for shipping low masses or
concentrations of fissile materials, and
licensees can submit a specific
exemption request under § 71.12 for
fissile materials that do not meet the
fissile exemption criteria in § 71.15.
The NRC received comments on
Issues 4 and 8 which suggested that the
NRC ‘‘grandfather’’ packages from
having to meet the revised
requirements. The NRC is proposing to
‘‘grandfather’’ older packages as
discussed in Issue 10, ‘‘Transitional
Arrangements.’’
Comments on Issue 4 on the proposed
insolation requirements stated that these
requirements would present challenges
to certificate holders, including cost to
certificate holders to evaluate the new
conditions; changing the units without
revising the corresponding values may
result in decreasing margins or
exceeding thermal limits; and the
insolation values are referenced in other
documents, which may have an impact
to the thermal evaluations for storage
systems certified under 10 CFR part 72.
While the NRC agrees there will be costs
with evaluating the new insolation
requirements, the NRC estimates that
the cost for existing certificates to show
compliance with the revised insolation
will be small, since the increased
insolation load would be approximately
3 percent. In addition, harmonizing
NRC requirements with those of IAEA
will ensure that packages approved by
the NRC would also be acceptable in
other countries where they might be
used for international transport. The
NRC made no changes as a result of this
comment. The NRC recognizes that all
packages age over time and that aging
effects should be considered for all
packages, not just for dual-purpose
packages.
The NRC received comments on Issue
9 opposing the addition of an aging
management program to 10 CFR part 71.
The commenters stated that, if such a
program were added, the program
should be limited to packages other than
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.
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One comment on Issue 10 supported
phasing out older packages as proposed
in transitional arrangements but
suggested a phase-out period longer
than 4 years. The NRC agreed and is
proposing an 8-year phase out of older
packages. As part of the NRC’s 2004
amendment to 10 CFR part 71 (69 FR
3697; January 26, 2004), certain
transportation packages, those
compatible with the 1967 edition of
Safety Series No. 6, became
unauthorized for use under the 10 CFR
part 71 general license after October 1,
2008. The NRC received requests to
extend the phase-out date beyond the
initial 4-year period to allow sufficient
time to design, obtain approval for, and
fabricate new packages. Given this
experience, in this proposed rule, the
NRC has selected a phase-out period of
8 years to give certificate holders
sufficient time to conduct these
activities, if needed. The NRC estimates
that it could take 2 to 4 years for design
of a new package and preparation of an
application, 1 to 2 years for package
approval, and 1 to 2 years for package
fabrication, depending on the package’s
complexity. Another comment on Issue
10 on transitional arrangements stated
that the NRC should not phase out
packages with a ‘‘–96’’ in the package
identification number and that the
proposed phase out of packages did not
consider the cost impact for designing
new packages. The NRC is not
proposing to phase out packages with a
‘‘–96’’ in the proposed rule, but rather
proposing to phase out packages that do
not have either a ‘‘–85’’ or a ‘‘–96’’ in
the package identification number (i.e.,
packages approved before April 1,
1996). The NRC included the cost of
designing a new package in the
regulatory analysis for the proposed
rule.
The NRC received one comment on
Issue 12 on the proposed quality
assurance program (QAP) changes,
stating that the proposed change would
be duplicative with 10 CFR part 50 QAP
requirements. The NRC disagrees with
this comment because if a 10 CFR part
50 licensee uses its 10 CFR part 50 QAP
for 10 CFR part 71 activities, the QAP
reporting requirements in 10 CFR part
50 would be controlling and 10 CFR
part 71 QAP reporting requirements
would not apply. Also, the NRC notes
that many users of 10 CFR part 71 do
not have 10 CFR part 50 licenses, and
the 10 CFR part 71 QAP change
provisions would not be duplicative for
them.
The NRC received a comment on
Issue 15 on the advance notification
requirements in § 71.97, stating that
there is no actual provision requiring
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advance notification for spent fuel
shipments. The requirements in § 71.97
currently contain reporting
requirements that are duplicative with
those in 10 CFR part 73, and the NRC
is proposing to delete the duplicative
language.
Because none of the comments would
result in significant changes to the draft
regulatory basis, the NRC considered
these comments in preparing this
proposed rule and did not issue a final
regulatory basis.
III. 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 Commissionissued 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):
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Issue
IAEA
harmonization
1
2
3
4.1
4.2
5
6
7
8
9
10
11
12
13
14
15.1
15.2
15.3
15.4
15.5
X
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Other
changes
No
action
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
X
Issue 1. Revision of Fissile Exemptions
The fissile material exemptions in
§ 71.15 and the fissile material general
licenses in §§ 71.22 and 71.23 allow
licensees to ship low-risk fissile
material (e.g., small quantities or low
concentrations) without meeting the
fissile material packaging requirements
and criticality safety assessments, as
specified in §§ 71.55 and 71.59, and
without obtaining prior NRC approval.
For these low-risk fissile material
shipments, the fissile material
exemptions and general licenses
provide reasonable assurance that
criticality safety is afforded under
normal conditions of transport and
hypothetical accident conditions. In
2012, IAEA modified the fissile
exception provisions in SSR–6,
paragraph 417, to include three new
per-package mass limit options, with
associated mass limits on the
consignment and/or conveyance.
The NRC proposes to incorporate two
additional fissile exemptions under
§ 71.15. This proposed rule would adopt
the exception in SSR–6, paragraph
417(c), without the associated
consignment limit of IAEA SSR–6,
paragraph 570(c). This proposed rule
would also adopt the exception in SSR–
6, paragraph 417(e), with its associated
exclusive use restriction in paragraph
570(e), but with a higher mass limit.
Since the amount of fissile material
allowed by SSR–6, paragraph 417(c), is
similar to the existing exemption in
§ 71.15(a), in terms of reactivity, the
NRC determined that the consignment
limit of IAEA SSR–6, paragraph 570(c),
is not necessary. Consignment limits, as
provided in 570(c), do not prevent the
accumulation of packages on a transport
conveyance, as there is no limit to the
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number of consignments that may be
present on a single conveyance.
Additionally, the number of these
packages does not need to be limited by
regulation because reaching the amount
required to approach criticality on a
single conveyance is not credible.
The NRC has determined that a mass
value higher than that contained in
IAEA SSR–6, paragraph 417(e), is
justified, given the conservatism
inherent in the exclusive use restriction
of the SSR–6 provision, and in basing
the mass limit on plutonium-239
(239Pu), which would have to be
shipped in a Type B package. The NRC
proposes a limit of 140 grams of fissile
material on a conveyance shipped under
exclusive use, as another exemption
under § 71.15. This limit is based on one
fifth of a minimum critical mass of
uranium-235 (235U) (as defined in
American National Standards Institute/
American Nuclear Society [ANSI/ANS]
8.1–2014 (Reaffirmed 2018), ‘‘Nuclear
Criticality Safety in Operations with
Fissionable Materials Outside
Reactors’’) under optimum conditions.
This mass represents a conservative
limit for fissile material, since five times
this amount would remain subcritical
under any condition. Additionally, the
limit provides safety equivalent to
packages approved under 10 CFR part
71 and could provide more flexibility
for shipping individual contaminated
items or small quantities of fissile
material. The NRC considers 235U for
this limit rather than 239Pu, as any
amount of 239Pu over 0.435 grams is
considered Type B, which would have
to be packaged to withstand both
normal and hypothetical accident
conditions of transport. Although the
NRC proposed value is different from
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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 § 71.15, only relieves the consignor
from having to ship in a ‘‘Fissile’’
package, evaluated per the requirements
of §§ 71.55 and 71.59. This material is
still subject to all other radioactive
materials transportation requirements in
10 CFR part 71 and in 49 CFR part 173
and should be packaged accordingly.
The NRC is proposing to make a minor
change to § 71.15(d) for clarity and to
maintain consistent language
throughout § 71.15.
Issue 2. Revision of Reduced External
Pressure Test for Normal Conditions of
Transport
The regulation at § 71.71(c)(3)
requires Type AF and Type B package
designs to be able to withstand a
reduction in external pressure to 25
kilopascals (kPa) (3.6 psia) under
normal conditions of transport. For a
Type A package (as defined in SSR–6,
paragraphs 231 and 429; 10 CFR 71.4,
‘‘Definitions’’; or 49 CFR 173.403,
‘‘Definitions’’), IAEA SSR–6, paragraph
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645, states that ‘‘[t]he containment
system shall retain its radioactive
contents under a reduction of ambient
pressure to 60 kPa.’’ This requirement
also applies to Type B(U) and Type
B(M) packages, in accordance with
SSR–6, paragraphs 652 and 667,
respectively. Additionally, IAEA SSR–6,
paragraph 621, indicates packages
containing radioactive material to be
transported by air shall be capable of
withstanding, without loss or dispersal
of the radioactive contents from the
containment system, an internal
pressure that produces a pressure
differential of not less than maximum
normal operating pressure plus 95 kPa
(13.8 psi).
In a final rule published by the DOT
(79 FR 40589; July 11, 2014), the DOT
harmonized its regulations in 49 CFR
chapter I to the 2009 Edition of IAEA
TS–R–1. In that final rule, the DOT
explained that a Type A package must
be designed to ensure the package can
retain its contents under the reduction
of ambient pressure. That ambient
pressure value, found at 49 CFR
173.412(f), was changed from 25 kPa
(3.6 psia) to 60 kPa (8.7 psia).
The NRC considered whether it
should change the reduced external
pressure test requirement in
§ 71.71(c)(3) to harmonize with the
IAEA transport standards and to be
consistent with the DOT regulations for
design requirements for Type A
packages. The NRC assessed the
potential impacts of the change in the
external pressure value from 25 kPa (3.6
psia) to 60 kPa (8.7 psia) and the
additional air transport requirements
from SSR–6, paragraph 621. The current
NRC reduced external pressure test
requirement, 25 kPa (3.6 psia), equates
to an altitude of about 35,000 feet
(10,668 meters) above sea level, which
is an appropriate altitude for air
transport of packages. Since cargo
planes use pressurized cargo holds
during air transport, this external
pressure value also represents the
ambient pressure on a package should
the cargo hold depressurize. Whereas
the 60 kPa (8.7 psia) value equates to an
altitude of about 14,040 feet (4,279
meters) above sea level. Thus, while the
60 kPa (8.7 psia) external pressure value
equates well with the highest paved
road in the United States (14,130 feet
(4,307 meters)) and with the elevation of
the highest operating freight railroad in
the United States (La Veta Pass at 9,242
feet (2,817 meters)), it would not
support air transport conditions, as
cargo planes operate at higher altitudes.
When comparing the current 25 kPa (3.6
psia) value with the proposed 60 kPa
(8.7 psia) value, and the associated
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altitudes, the NRC determined that no
change to § 71.71(c)(3) is needed, and
the 25 kPa (3.6 psia) value should be
retained.
The NRC also considered adding the
air transport requirements from SSR–6,
paragraph 621. However, other than
specific air transport requirements at
§ 71.55(f), ‘‘General requirements for
fissile material packages’’ and § 71.88,
‘‘Air transport of plutonium,’’ 10 CFR
part 71 does not contain ‘‘modespecific’’ regulations. Because the
existing reduced external pressure test
value covers air transport conditions as
discussed above, and because of the
robustness of Type AF and Type B
packages, as compared to Type A
packages, the NRC finds it unnecessary
to add the mode-specific air transport
requirements from SSR–6, paragraph
621, into 10 CFR part 71.
Based on the above considerations
and assessments, the NRC has decided
not to pursue any changes to
§ 71.71(c)(3). As a result, no further
discussion or analysis is presented in
this proposed rule on the reduced
external pressure test for normal
conditions of transport.
Issue 3. Inclusion of Type C Package
Standards
In the 2004 final rule, the NRC did not
adopt the regulations for Type C
packages contained in IAEA TS–R–1.
The NRC did not adopt them because 1)
§§ 71.64 and 71.74 for plutonium air
transportation contain more rigorous
packaging standards, 2) the NRC
perceived no need (current or
anticipated) for such packages, and 3) if
a need arose for import or export, it
could be accomplished through the DOT
regulations.
In the request for comment on the
issues paper, the NRC asked
stakeholders whether there was a need
for domestic transport of Type C
packages. No NRC licensees expressed a
need for domestic transport of Type C
packages. Therefore, the NRC has
decided not to pursue further changes to
Type C package standards as
contemplated in the regulatory basis
document. As a result, no further
discussion or analysis is presented in
this proposed rule on that issue.
Issue 4. Revision of Insolation
Requirements for Package Evaluations
During transport, a package is
subjected to heating by the sun, called
insolation. The effect of insolation is an
increase in the package temperature.
The NRC is proposing to change the unit
of measure for the values of insolation
used for the heat test for normal
conditions of transport in § 71.71(c)(1),
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and to add insolation to the initial
conditions for the tests for hypothetical
accident conditions in § 71.73(b).
Issue 4.1. Revision of Units for
Insolation for Normal Conditions of
Transport
The units for insolation in 10 CFR
part 71 are gram calories per square
centimeter (g cal/cm2). When the IAEA
published Safety Series No. 6,
‘‘Regulations for the Safe Transport of
Radioactive Material, 1985 Edition,’’ it
revised the units used for insolation for
normal conditions of transport from a
hybrid of English and metric units (g
cal/cm2) to metric units (watts per
square meter (W/m2)). When the IAEA
changed the units, it chose to keep the
same numerical values, thus increasing
the evaluated solar heat load on a
package by approximately 3 percent.
The IAEA did not provide a technical
rationale for this change; however, the
NRC observes that retaining the existing
numerical quantities maintains simple
(round) values in the regulations that
result in a small change in solar heat
load.
The NRC previously harmonized its
regulations with the 1985 Edition of
Safety Series No. 6 (60 FR 50248;
September 28, 1995). That final rule
neither discussed nor proposed
changing the units on the heat test for
normal conditions of transport in
§ 71.71(c)(1). Consequently, the current
units for insolation in 10 CFR part 71
are ‘‘g cal/cm2.’’ This is inconsistent
with IAEA standards in the 2018
Edition of SSR–6. As a result, NRC
package approvals are evaluated for less
insolation than that prescribed by IAEA
standards and evaluated for approval by
foreign competent authorities.
The NRC is proposing to revise the
units of insolation for the heat test for
normal conditions of transport in
§ 71.71(c)(1) to match the units used in
the 2018 Edition of SSR–6 to ensure that
NRC requirements for insolation are
consistent with the IAEA standard.
Consistent with Issue 10, ‘‘Transitional
Arrangements,’’ the NRC would not
expect a certificate holder to evaluate
the higher solar heat load unless it
requests a revision of its certificate to
show compliance with the revised
transportation regulations in 10 CFR
part 71. Additionally, given the small
increase in insolation due to the revised
units, the NRC expects that certificate
holders will be able to show compliance
with the package approval standards in
subpart E, ‘‘Package Approval
Standards,’’ to 10 CFR part 71.
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Issue 4.2. Inclusion of Insolation for
Hypothetical Accident Conditions
In Safety Series No. 6, ‘‘Regulations
for the Safe Transport of Radioactive
Material, 1985 Edition (As Amended
1990),’’ paragraph 628 stated, ‘‘With
respect to the initial conditions for the
thermal test, the demonstration of
compliance shall be based upon the
assumption that the package is in
equilibrium at an ambient temperature
of 38 °C. The effects of solar radiation
may be neglected prior to and during
the tests, but must be taken into account
in the subsequent evaluation of the
package response.’’
The thermal test, previously in
paragraph 628, was moved to paragraph
728 in the 1996 Edition of TS–R–1 and
revised to state, ‘‘The specimen shall be
in thermal equilibrium under conditions
of an ambient temperature of 38 °C,
subject to the solar insolation conditions
specified in Table XI and subject to the
design maximum rate of internal heat
generation within the package from the
radioactive contents.’’
When the NRC revised its regulations
in 2004 to harmonize with the 1996
IAEA standards (69 FR 3697; January
26, 2004), the NRC did not revise the
initial conditions of the fire test listed
in § 71.73(b) to require evaluation of
insolation as an initial condition.
Since a fire can occur on a hot, sunny
day, and to be consistent with IAEA
standards, the NRC is proposing to
revise the initial conditions in § 71.73(b)
to require insolation as an initial
condition for all the tests for
hypothetical accident conditions.
Consistent with Issue 10, ‘‘Transitional
Arrangements,’’ the NRC would expect
a certificate holder to evaluate the
revised initial conditions in § 71.73 if it
wants to revise its certificate to show
compliance with the revised
transportation regulations in 10 CFR
part 71.
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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
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the current regulations in 10 CFR part
71 use the term ‘‘radiation level,’’ the
NRC is concerned that using a different
term from the IAEA to define external
radiation standards could create some
confusion with respect to international
shipments.
Additionally, NRC regulations in 10
CFR part 20, ‘‘Standards for Protection
Against Radiation,’’ include a definition
for ‘‘dose equivalent’’ in § 20.1003 that
means the product of the absorbed dose
in tissue, quality factor, and all other
necessary modifying factors at the
location of interest. The units of dose
equivalent are the rem and sievert (Sv).
The NRC considered replacing the
term ‘‘radiation level’’ used throughout
10 CFR part 71 with ‘‘dose equivalent
rate.’’ However, this change would
result in cost impacts to licensees to
change documentation and training
programs with no safety benefit.
Therefore, in order to minimize the
burden to licensees, the NRC is
proposing to add a definition to § 71.4
that clarifies that ‘‘radiation level’’
means ‘‘dose equivalent rate,’’ which
enables the NRC to continue using
‘‘radiation level’’ throughout 10 CFR
part 71. The NRC is not expecting any
licensee to change its documentation to
account for this new definition.
Issue 6. Deletion of Low Specific
Activity-III Leaching Test
The definition for ‘‘Low Specific
Activity (LSA) material’’ in § 71.4
includes three categories of material:
LSA–I, LSA–II, and LSA–III.
Radioactive material, low specific
activity category III (i.e., LSA–III)
includes solids, excluding powders, that
meet the requirements in § 71.77,
‘‘Qualification of LSA–III material’’ and
which have an estimated average
specific activity limit that does not
exceed 2 × 10¥3 times the A2 value per
gram (A2/g). The qualification tests in
§ 71.77 include a leaching test with
immersion of the specimen material for
7 days. The IAEA eliminated the LSA–
III leaching test in SSR–6, 2018 Edition,
from paragraphs 409, 601, and 701.
Consequently, the NRC is proposing
corresponding revisions to §§ 71.4,
71.77, and 71.100, ‘‘Criminal penalties,’’
to remove the leaching test and its
references.
In April 2015, an international
working group meeting was conducted
to discuss issues related to LSA–II and
LSA–III material, with special attention
on the need for the LSA–III leaching
test. The need for the leaching test was
questioned because the working group
determined that the test has no bearing
on the inhalation risk of exposure to
material during transport. The
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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 × 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 nondispersible 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 20fold increase in average specific activity
between LSA–III and LSA–II material.
The NRC recognizes the working
group’s information, and is
recommending harmonization with
SSR–6, 2018 Edition, and removal of the
leaching test from 10 CFR part 71. The
NRC agrees that requiring the LSA–III
leaching test does not increase the safety
of the material during transport.
Further, the test does not decrease the
inhalation pathway exposure when
compared to LSA–II material in powder
form, and therefore should be removed
from 10 CFR part 71. The NRC
considered the information provided by
the LSA–II and LSA–III working groups
and comments received on this issue
during the comment period on the
NRC’s issues paper. Additionally, the
NRC considers that removal of the
leaching test also would reduce
regulatory burden for shippers, while
still maintaining reasonable assurance
of safety for transport of LSA–III
material.
The NRC is proposing to remove the
leaching test in § 71.77 and make
conforming changes to §§ 71.4 and
71.100, which both reference § 71.77.
Issue 7. Inclusion of New Definition for
Surface Contaminated Object
As more nuclear facilities begin
decommissioning activities, there will
be an increase in the number of
shipments of radioactive materials from
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these facilities. Decommissioning
activities can include transporting large
radioactive objects (e.g., steam
generators, coolant pumps, and
pressurizers). Under current NRC
regulations, shipment of such large,
nonstandard packages that do not meet
the existing definition of surface
contaminated objects (i.e., either SCO–
I or SCO–II, as defined in § 71.4) could
be addressed through a special package
authorization under § 71.41(d).
However, such an authorization may
take significant time. The NRC proposes
to add a regulatory definition for SCO–
III to include these types of objects,
allowing a shipper to more
appropriately categorize the item it is
planning to transport. The NRC
anticipates an increase in efficiency for
both the NRC and licensees when the
SCO–III definition is included in 10
CFR part 71 when compared to the
special package authorization review
needed under § 71.41(d). Harmonization
with SSR–6, 2018 Edition, would add
the new SCO–III category and the
associated definition.
In the 2004 final rule (69 FR 3697;
January 26, 2004), the NRC determined
that special package authorizations were
necessary because there were no
regulatory provisions in 10 CFR part 71
concerning large, nonstandard packages
considered for transportation. Therefore,
the NRC added paragraph (d) to § 71.41.
Since that time, the NRC has gained
experience with the safety aspects of
shipping these types of large, nonstandard packages. For example, in
2006, the LaCrosse reactor vessel was
the first shipment in which a package
was approved under § 71.41(d). In
addition, a special package
authorization was issued for the West
Valley Melter Package from the West
Valley Demonstration Project. In the
future, a licensee shipping large
radioactive objects that have been
determined to meet the definition of
SCO–III would not need NRC review
and approval for a special package
authorization.
Both the NRC and DOT intend to add
a definition for SCO–III. The NRC is
coordinating with the DOT to align its
definition with the DOT’s, since the
DOT is the lead agency for review and
evaluation of both LSA and SCO
material.
Issue 8. Revision of Uranium
Hexafluoride Package Requirements
In the 2004 final rule (69 FR 3697;
January 26, 2004), the NRC harmonized
its regulations with the 1996 Edition of
IAEA TS–R–1. In that final rule, the
NRC added a new provision, § 71.55(g),
to provide a specific exception for
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certain uranium hexafluoride (UF6)
packages from the requirements of
§ 71.55(b). The exception allows UF6
packages to be evaluated for criticality
safety without considering inleakage of
water into the containment system,
provided certain conditions are met,
including that the uranium is enriched
to not more than 5 weight percent in
235U. To use this exception, the
applicant must demonstrate, among
other things, that, following the tests for
hypothetical accident conditions in
§ 71.73, there is no physical contact
between the valve body and any other
component of the packaging, other than
at its original point of attachment, and
the valve remains leak tight. ‘‘Leaktight’’
is defined in ANSI N14.5–2014,
‘‘American National Standard for
Radioactive Materials—Leakage Tests
on Packages for Shipment,’’ as ‘‘[t]he
degree of package containment that, in
a practical sense, precludes any
significant release of radioactive
materials. This degree of containment is
achieved by demonstration of a leakage
rate less than or equal to 1 × 10¥7
ref·cm3/s, of air at an upstream pressure
of 1 atmosphere (atm) absolute (abs),
and a downstream pressure of 0.01 atm
abs or less.’’
The NRC provided the specific
exception: (1) to be consistent with the
worldwide practice and limits
established in national and international
standards (ANSI N14.1–2012, ‘‘Nuclear
Materials—Uranium Hexafluoride—
Packagings for Transport,’’ and
International Organization for
Standardization 7195, ‘‘Packaging of
Uranium Hexafluoride (UF6) for
Transport’’) and DOT regulations (49
CFR 173.417(b)(5)); (2) because of the
history of safe shipment; and (3)
because of the essential need to
transport the commodity. In that final
rule, the NRC codified its long-standing
practice to not consider water inleakage
into UF6 packages as long as the
documentation of the results of the tests
for hypothetical accident conditions
tests at § 71.73 show that the cylinder
valve was not affected.
In SSR–6, 2018 Edition, the IAEA
added the same standard for the plug as
was added in the 1996 Edition of TS–
R–1 for the valve to ensure that the
entire cylinder remains leak tight. The
revised paragraph 680(b)(i), SSR–6,
2018 Edition, states: ‘‘Packages where,
following the tests prescribed in para.
685(b), there is no physical contact
between the valve or the plug and any
other component of the packaging other
than at its original point of attachment
and where, in addition, following the
test prescribed in para. 728, the valve
and the plug remain leaktight.’’
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The 30-inch UF6 cylinder, the most
commonly used cylinder to transport
large quantities of enriched UF6 for the
fuel fabrication industry, has two
penetrations: one for the valve at the top
to fill the cylinder and one for the drain
plug at the bottom used during
maintenance. In order to ensure
criticality safety, both the plug and the
valve must remain leak tight after the
tests for hypothetical accident
conditions to prevent ingress of water
into the cylinder. While this may be a
new requirement in transportation
regulations, during package approval,
the NRC has always verified that the
entire 30B cylinder remained leak tight
after the tests for hypothetical accident
conditions.
The NRC is proposing to revise
§ 71.55(g)(1) to require that there is no
contact between the cylinder plug and
any other part of the packaging, other
than at its original attachment point and
that the cylinder plug remains leak
tight, as NRC requires for the cylinder
valve.
Issue 9. Inclusion of Evaluation of Aging
Mechanisms and a Maintenance
Program
The NRC regulations do not explicitly
require that a package application
include an evaluation of aging
mechanisms and a maintenance
program. Rather, applicants include an
evaluation of aging effects on package
components to ensure there is no
significant degradation in accordance
with § 71.43(d). The NRC regulations at
§ 71.43(d) require that packages be made
of materials and construction that assure
that there will be no significant
chemical, galvanic, or other reaction
(including effects of irradiation from the
package contents) among the packaging
components, among package contents,
or between the packaging components
and the package contents, including
possible reaction resulting from
inleakage of water, to the maximum
credible extent.
For those components where aging is
detrimental to package performance,
applicants provide a description of the
maintenance program, including
periodic testing to evaluate the
components’ efficacy and/or a
replacement or repair schedule, to
mitigate those detrimental effects. The
NRC requires that licensees and CoC
holders follow the maintenance
program, which is provided in the
application for approval, as a condition
of approval in the CoC. Additionally,
NRC regulations at § 71.87(b) require
that, prior to each shipment, the
licensee ensures that the package is in
unimpaired physical condition except
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for superficial defects such as marks or
dents. Meeting this regulation, along
with the scheduled periodic tests and
replacement/repair in the maintenance
program, should identify package
deterioration prior to age-related
degradation becoming a safety issue
during transport.
In paragraph 613A, SSR–6, 2018
Edition, the IAEA added that package
design evaluations must consider aging
mechanisms. In paragraph 809, SSR–6,
2018 Edition, the IAEA added that the
application for package approval must
contain a maintenance program.
Because an evaluation of aging effects
and a description of the maintenance
program are not specifically required by
10 CFR part 71, the NRC is proposing
to revise § 71.43(d) to specifically
include the evaluation of the effects of
aging, and add a new provision to
subpart D, ‘‘Application for Package
Approval,’’ to include a description of
the maintenance program in an
application for package approval, to
better align with these standards in
SSR–6, 2018 Edition.
Issue 10. Revision of Transitional
Arrangements
Historically, IAEA standards and DOT
and NRC regulations have included
transitional arrangements when the
regulations have undergone revision.
The purpose is to minimize the costs
and impacts of implementing changes in
the regulations, since package designs
and special form sources that are
compliant with the existing regulations
do not become unsafe when the
regulations are revised (unless a
significant safety issue is corrected in
the revision).
Typically, the transitional
arrangements include provisions that
allow for (1) continued use of existing
package designs and packagings already
fabricated; and completion of
packagings in the process of being
fabricated, although some restrictions
on fabrication of packagings approved to
earlier editions of the regulations may
be imposed; (2) restriction on
modifications to package designs
without the need to demonstrate full
compliance with the revised
regulations; (3) changes in packaging
identification numbers; and (4) changes
to the fabrication and use of special
form sources approved to earlier
versions of the regulations.
The NRC CoCs include a package
identification number which identifies
the NRC regulations and the
corresponding version of IAEA
standards to which the package was
approved. For example, packages with a
‘‘–85’’ in the package identification
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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., NRCapproved 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
§ 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
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some additional restrictions on approval
and fabrication. The IAEA’s SSR–6,
2018 Edition, prohibits fabrication of
special form radioactive material that
received unilateral approval under the
1985 Edition of Safety Series No. 6 or
1985 (as Amended 1990) Edition of
Safety Series No. 6. Also, after
December 31, 2025, IAEA standards
prohibit new fabrication of special form
radioactive material sources to a design
that had received unilateral approval
under the 1996 Edition; 1996 Edition
(Revised); 1996 (as Amended 2003)
Edition of TS–R–1; TS–R–1, 2005
Edition; TS–R–1, 2009 Edition; and
SSR–6, 2012 Edition.
Finally, in paragraphs 832–833, SSR–
6, 2018 Edition, the IAEA revised the
package identification number in the
CoC to delete the year designation (i.e.,
‘‘–85’’ or ‘‘–96’’) for those package
designs that are approved to SSR–6,
2018 Edition.
In the 2004 final rule (69 FR 3698;
January 26, 2004), the NRC adopted the
following grandfathering provisions in
§ 71.19 for previously-approved
packages:
• 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 § 71.4 because NRC
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regulations were already consistent with
the 1996 Edition of TS–R–1.
The NRC rulemaking in 2015 (80 FR
33988; June 12, 2015) made two minor
changes to the transitional arrangements
regulations. First, the grandfathering
provision that was in § 71.19(a) for
packages approved under NRC
standards that were compatible with the
provisions of the 1967 Edition of Safety
Series No. 6 was deleted since that
provision expired on October 1, 2008.
Second, the definition of ‘‘special form
radioactive material’’ was revised to
allow special form radioactive material
that was successfully tested using the
current requirements of § 71.75(d) to
continue to qualify as special form
radioactive material, if the testing was
completed before September 10, 2015.
Consistent with past practices, the
NRC is proposing transitional
arrangements to phase out older
packages without a ‘‘–85’’ or ‘‘–96’’ in
the package identification number, and
limit use of packages with a ‘‘–96’’ to
those whose fabrication has been
completed by December 31, 2028, and
consistent with DOT, limit fabrication of
special form sources. The NRC
determined that it is appropriate to
begin a phased discontinuance of these
older packages to further harmonize
NRC’s regulations with the IAEA
standards in 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
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packaging was completed after
December 31, 2028, and require
multilateral approval (as defined in 49
CFR 173.403, ‘‘Definitions’’) for
packages to be used for international
shipment after December 31, 2025.
Revise § 71.17(e) to state that packages
with a ‘‘–96’’ in the package
identification number would become
previously approved packages and
subject to the current § 71.19(c).
3. Coordinate with the DOT and make
appropriate changes to § 71.4 to align
with the definition of ‘‘special form
radioactive material’’ that the DOT is
proposing to adopt as part of their
harmonization rulemaking, since DOT is
the lead for certifying special form
sources. The NRC is proposing to allow
continued use of special form
radioactive material that was approved
to the regulations in effect from October
1, 2004 to the effective date of this
rulemaking, provided they are
fabricated on or before December 31,
2025.
4. Allow for package designs with a
‘‘–96’’ or earlier package identification
number to be resubmitted to the NRC for
review against the current standards. If
the package design described in the
resubmitted application meets the
current standards, the NRC may issue a
new CoC for that package design
without a year designation.
The NRC notes that the IAEA
eliminated the approval year in the
package identification number for
packages approved to SSR–6, 2018
Edition. Packages that were approved to
NRC regulations harmonized with the
1973 Edition of Safety Series No. 6 do
not have a year designation in the
package identification number. To avoid
confusion regarding these older
packages, the NRC would revise all
existing CoCs that do not have a ‘‘–85’’
or ‘‘–96’’ in their package identification
number to add a provision that those
CoCs cannot be renewed beyond the end
date of the 8-year phase out period
without being recertified to the revised
version of 10 CFR part 71.
Issue 11. Inclusion of Head Space for
Liquid Expansion
The NRC’s regulation in § 71.87,
‘‘Routine determinations,’’ requires that
before each shipment of licensed
material, the licensee must ensure that
the package, which includes its
contents, satisfies the applicable
requirements of part 71. One such
requirement is that the licensee must
determine in accordance with § 71.87(d)
that any system for containing liquid is
adequately sealed and has adequate
space or other specified provision for
expansion of the liquid.
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The NRC’s requirement in § 71.87(d)
is compatible with the DOT’s
regulations at 49 CFR 173.24(h)(1),
‘‘General requirements for packagings
and packages.’’ That regulation requires:
‘‘When filling packagings and
receptacles for liquids, sufficient ullage
(outage) must be left to ensure that
neither leakage nor permanent
distortion of the packaging or receptacle
will occur as a result of an expansion of
the liquid caused by temperatures likely
to be encountered during
transportation.’’
The DOT’s regulations in 49 CFR
173.412(k), ‘‘Additional design
requirements for Type A packages,’’
contain a general design requirement for
Type A packages designed to contain
liquids to ensure that packages provide
for ullage to accommodate variations in
temperature of the contents. The term
‘‘ullage’’ refers to the unfilled space in
a container, or the amount by which the
contents of a container fall short of
being full. Because DOT’s regulations
for Type AF, Type B, and Type BF
packages refer to the NRC’s regulations,
DOT’s regulations do not contain design
requirements for Type AF, Type B, or
Type BF packages. Type A, Type AF,
Type B, and Type BF packages are
defined in § 71.4, ‘‘Packages.’’
The IAEA standards in paragraph 649,
SSR–6, 2018 Edition, require that ‘‘The
design of a package intended for liquid
radioactive material shall make
provision for ullage to accommodate
variations in the temperature of the
contents, dynamic effects and filling
dynamics.’’
The NRC regulations have an
operational requirement in § 71.87(d) to
ensure that for a system containing
liquid, there is sufficient head space, or
other specified provision to
accommodate the expansion of liquid.
The NRC does not, however, have a
comparable design requirement for Type
AF and Type B packages in 10 CFR part
71 to that in DOT’s regulations. Even
though the NRC’s regulations do not
include a comparable design
requirement for ensuring sufficient
space to allow for liquid expansion, any
Type AF or Type B package design
certified by the NRC must comply with
§ 71.87 and DOT regulations in 49 CFR
173.24(h) on ullage when being filled.
During review of applications for
either a new CoC or an amendment to
an existing CoC, the NRC reviews
whether the requirements in § 71.87(d)
are reflected in the operating procedures
for packages with liquid contents. Each
package approval issued by the NRC
contains a condition to ensure that the
package is prepared in accordance with
the operating procedures in the
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On June 12, 2015, the NRC issued a
final rule (80 FR 33988), updating the
administrative procedures for the QAP
requirements described in 10 CFR part
71, subpart H, ‘‘Quality Assurance.’’
Specifically, the NRC added § 71.106 to
establish requirements for QAP changes
and associated reporting requirements.
Previously, all changes made to QAP
approvals had to be reviewed and
approved by the NRC before they could
be implemented. The provisions in
§ 71.106 allow changes to QAPs that do
not reduce commitments, such as those
that involve administrative
improvements and clarifications,
spelling corrections, and nonsubstantive changes, to be made and
implemented without prior NRC
_
CSI - 10
where X, Y, and Z are mass limits of 235U,
233U, and plutonium obtained from
Table 71–1 (if 233U or plutonium are
present) or Table 71–2.
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Similarly, the general license criteria
in § 71.23 allow NRC licensees to ship
small quantities of special form
plutonium in packages that have been
assigned a CSI to ensure accumulation
[grams of 235 U
X
_
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grams of 233 U
y
+
[grams of
239 Pu+
grams of
24
reflection, i.e., optimally moderated
spheres of 235U, 233U, and 239Pu with
full water reflection. The mass limits in
§ 71.23 have a similar basis, but are
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Issue 13. Deletion of Type A Package
Limitations in Fissile Material General
Licenses
The general license criteria in § 71.22
allow NRC licensees to ship small
quantities of fissile material in packages
that have been assigned a criticality
safety index (CSI) to ensure
accumulation control for packages on a
conveyance. The provisions of § 71.22
require that (1) the fissile material is in
a Type A package that meets the
requirements of 49 CFR 173.417(a); (2)
licensees have an NRC-approved QAP
satisfying the provisions of 10 CFR part
71, subpart H; (3) there is no more than
a Type A quantity of radioactive
material; (4) there is less than 500 grams
total of beryllium, graphite, or
hydrogenous material enriched in
deuterium; and (5) the package is
labeled with a CSI that meets the limits
in § 71.22(d). The regulation in
§ 71.22(e)(1) provides an equation to
calculate package CSI:
grams of Pu]
z
control for packages on a conveyance.
The provisions of § 71.23 require that (1)
the fissile material is in a Type A
package meeting the requirements of 49
CFR 173.417(a); (2) licensees have an
NRC-approved quality assurance
program satisfying the provisions of 10
CFR part 71, subpart H; (3) there is no
more than a Type A quantity of
CS/ - 10
The calculations that support the
mass limits in § 71.22 include
conservative assumptions regarding
neutron moderation and water
+
reporting requirements in 10 CFR part
71 would be consistent with those in
§§ 50.54(a)(3) and 50.71(e)(2) for 10 CFR
part 50 QAPs. Since the 2015 final rule
became effective, the NRC has received
questions and concerns from industry
on this subject since the language in
§ 71.106 does not state that QAP
approval holders must report even if
there were no changes in the prior 24month period.
The NRC is proposing to revise
§ 71.106(b) to clarify that a biennial
report must be submitted to the NRC
even if no changes are made to the QAP
during the reporting period.
radioactive material; (4) there is less
than 1,000 grams of plutonium,
provided that the total amount of 239Pu
and 241Pu constitutes less than 240
grams of the plutonium in the package;
and (5) the package is labeled with a CSI
that meets the limits in § 71.23(d). The
regulation in § 71.23(e)(1) provides an
equation to calculate package CSI:
241
Pu]
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
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Issue 12. Revision of Quality Assurance
Program Biennial Reporting
Requirements
approval. QAP changes that would
reduce commitments require prior NRC
approval.
In addition, § 71.106 requires that
changes to QAPs that do not reduce
commitments must be submitted to the
NRC every 24 months. That final rule
also specified, ‘‘If a quality assurance
program approval holder has not made
any changes to its approved quality
assurance program description during
the preceding 24-month period, the
approval holder will be required to
report this to the NRC’’ (80 FR 33994).
In addition, the NRC’s guidance
document for 10 CFR part 71 QAPs,
Regulatory Guide 7.10, Revision 3, was
updated in conjunction with the 2015
final rule to state that if no changes were
made to the QAP, a QAP approval
holder would indicate to the NRC that
no changes were made.
The requirement for a report, even if
no changes were made during the
preceding 24-month period, is necessary
as the NRC inspection program for 10
CFR part 71 QAP approval holders
relies on having current information
about the QAP available to the NRC.
The NRC considers the 24-month
reporting requirement, including when
no changes are made, as providing an
appropriate balance between the burden
placed on the QAP approval holders
and the need to ensure that the NRC has
current information for its oversight of
these QAPs. Most QAP approval holders
subject to periodic inspection are
inspected every 5 years or on an asneeded basis. Another benefit to
receiving a report even when no QAP
changes have been made is that the QAP
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application. This ensures that all
package users, whether NRC licensees
or not, comply with the requirements
listed in § 71.87, as appropriate for the
package design.
Although the NRC regulations ensure
that adequate ullage exists, the NRC has
received on occasion an application that
did not evaluate whether there was
sufficient design space in a container
with liquids. To clarify this
requirement, the NRC is proposing to
revise § 71.43, ‘‘General standards for all
packages,’’ to add a design requirement
for a package designed to contain
liquids to ensure adequate ullage during
evaluation of the tests and conditions
for normal conditions of transport and
hypothetical accident conditions.
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material will remain in the package
under normal conditions of transport
because of the Type A package
requirement but can reconfigure outside
of the package under hypothetical
accident conditions. The limitation to a
Type A quantity of radioactive material
in a Type A package, however, is not
consistent with the mass limits for some
fissile nuclides in some cases (e.g., the
mass limits for 239Pu in Table 71–1 are
37 grams or 24 grams, depending on the
degree of moderation, while the A2
value for 239Pu is equivalent to 0.435
grams). In addition, the requirement in
§ 71.23 does not consistently refer to
‘‘special form sealed sources’’ in that
paragraph (a) also refers to 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 §§ 71.22 and 71.23 in a Type
B package. When shipping material that
meets the mass limits of the general
licenses in §§ 71.22 and 71.23 in a Type
B package, the criticality safety
conclusions associated with these mass
limits remain valid. In fact, the material
would be less likely to present a
criticality hazard, as Type B packages
generally are more robust and have
more mass, which would increase
neutron absorption, limit releases under
hypothetical accident conditions, and
prevent material from multiple packages
from redistributing together under
optimum moderation conditions.
Revising the general licenses to
authorize transport in a Type B package
would also require conforming changes
to § 71.0(d)(1). The regulations in
§ 71.0(d)(1) state that use of the general
licenses in § 71.22 or § 71.23 does not
require NRC approval. Package approval
is not currently required by the NRC
because the conditions of the general
licenses require the contents to be in a
Type A package. The regulations in
§ 71.14(b)(1) exempt the licensee from
all requirements in 10 CFR part 71,
except for §§ 71.5 and 71.88, when
shipping a Type A quantity. Because the
NRC is proposing to revise §§ 71.22 and
71.23 to authorize shipment of a Type
B quantity of radioactive material, an
NRC package approval would be
required for shipment of the Type B
quantity of radioactive material. The
NRC package approval for the Type B
quantity of radioactive material would
not include evaluation of criticality
safety because the criticality safety is
assured for shipment of fissile material
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authorized under one of these general
licenses.
While NRC is not proposing to revise
§§ 71.22(b) and 71.23(b), which require
that the licensee have an NRC-approved
QAP. Applications for QAP approvals
use a graded approach, based on the
planned activities and shipments that a
licensee plans to make. For example, if
a licensee has a QAP that was approved
for making only Type A shipments
under § 71.22 or § 71.23, then the
licensee would need to obtain
additional NRC approval for a QAP that
includes QA items necessary for making
Type B shipments.
In addition, because the NRC is
proposing to authorize shipments of
Type B packages in §§ 71.22 and 71.23,
the NRC is proposing to include three
new paragraphs in §§ 71.22 and 71.23
that are similar to the requirements in
§ 71.17(c), (d), and (e). The NRC is
proposing to add a new requirement in
§§ 71.22(f) and 71.23(f) to ensure that,
for shipments made using the respective
general license, each licensee must
comply with § 71.17(c), i.e., the licensee
must: (1) maintain a copy of the NRC
approval, including all referenced
documents; (2) comply with the terms
and conditions of the NRC approval and
the applicable requirements of subparts
A, G, and H in 10 CFR part 71; and (3)
prior to first use, register to use the
package. A licensee is only required to
register once to use a package, and
therefore a licensee already registered to
use the package via § 71.17 would not
have to re-register to use the package
under one of these two general licenses.
The NRC is proposing to add a new
requirement in §§ 71.22(g) and 71.23(g)
to state that, for a package to be used
under the respective general license, the
NRC package approval must state that
the package can be used under the
general license in either § 71.17 or the
general license in § 71.22 or § 71.23.
Authorizing use under the general
license in § 71.17 would ensure that
existing, approved Type B package
designs could also be used to transport
the material authorized by one of the
two general licenses in § 71.22 or
§ 71.23.
Finally, the NRC is proposing to add
a new requirement in §§ 71.22(h) and
71.23(h) to ensure that any Type B
package used under the respective
general license approved by the NRC
before the effective date of the final rule
is subject to the transitional
arrangements in § 71.19. Issue 10 in
Section III of this document describes
the NRC’s proposed changes to its
transitional arrangements.
In summary, the NRC is proposing to
remove the restriction in §§ 71.22 and
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71.23 to ship Type A material in only
a Type A package (i.e., allowing
shipment of material up to the mass
limits in a Type B package); to add three
new paragraphs in §§ 71.22 and 71.23;
and to make conforming changes to
§ 71.0(d)(1). Additionally, the NRC is
proposing to clarify that only special
form sealed sources, not just sealed
sources may be delivered to a carrier for
transport using the general license in
§ 71.23.
Issue 14. Deletion of 233 U Restriction in
Fissile General License
The general license criteria in § 71.22
allow NRC licensees to ship small
quantities of fissile material in packages
that have been assigned a CSI to ensure
accumulation control for packages on a
conveyance. General license users
assign a CSI based on the equation in
§ 71.22(e)(1), and the fissile mass limits
in either Table 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 235U.
However, § 71.22(e)(5) states in part that
the lower mass values of Table 71–1
must be used if the enrichment level of
uranium is unknown, if the amount of
plutonium exceeds one percent of the
mass of 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
§ 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 § 71.22, without affecting
criticality safety.
The NRC is therefore proposing to
revise § 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 § 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
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clarifications to Agreement State
regulations.
Issue 15.1. Deletion of Duplicative
Reporting Requirements
In the 2002 proposed rule (67 FR
21390, April 30, 2002), the NRC
proposed changes to its reporting
requirements in § 71.95, ‘‘Reports.’’
Those proposed changes would have:
(1) required licensees to obtain
certificate holder input before
submitting an event report; (2) provided
direction on the content of the written
report; and (3) lengthened the reporting
requirement date to 60 days, consistent
with other reporting requirements in
NRC regulations. The proposed rule
recommended adding 71.95(a)(1) and (2)
and 71.95(b), but not the current
71.95(a)(3).
In the final rule (69 FR 3697, January
26, 2004), the NRC stated that the
proposed rule had inadvertently left out
new paragraph (a)(3), mentioned in the
proposed rule’s regulatory analysis, that
would retain the existing requirement
for licensees to report instances of
failure to follow the conditions of the
CoC while a packaging was in use.
Paragraph (a)(3) was thus added to the
final rule. However, in adding that
paragraph to the final rule, the NRC
introduced duplicative language
between it and paragraph (b).
The NRC is proposing to delete the
duplicative text in paragraph (a)(3).
lotter on DSK11XQN23PROD with PROPOSALS1
Issue 15.2. Revision of the Definition of
Low Specific Activity
The NRC is proposing to modify the
first sentence in the definition of ‘‘Low
Specific Activity (LSA) material’’ in
§ 71.4 to change ‘‘excepted under
§ 71.15’’ to ‘‘exempted under § 71.15.’’
This change would make the definition
of LSA in § 71.4 consistent with the title
of § 71.15, ‘‘Exemption from
classification as fissile material’’ and
ensure that it is clear that LSA packages
may contain fissile material up to the
exemption limits in § 71.15.
Issue 15.3. Revision of Tables
Containing A1 and A2 Values and
Exempt Material Activity and
Consignment Limits
The IAEA has made changes in SSR–
6, 2018 Edition, related to the A1 and A2
activity values and the exempt material
activity concentrations and exempt
consignment activity limits. The DOT is
the lead agency for information related
to the A1 and A2 values and for the
exempt material activity concentrations
and exempt consignment activity limits,
as provided in 49 CFR 173.435 and
173.436, respectively. The NRC has
corresponding information in 10 CFR
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part 71, Appendix A, Tables A–1 and
A–2.
To be considered radioactive material
under DOT’s regulations (i.e., Class 7
(radioactive) material as defined in 49
CFR 173.403), the material must exceed
both the nuclide specific exemption
concentration limit and the
consignment exemption activity limit.
The A1 and A2 values are quantities of
radioactivity that are used in the
transportation regulations to determine
the type of packaging necessary for a
particular radioactive material
shipment. Each radionuclide is assigned
an A1 and an A2 value, where A1 is the
maximum activity of special form
material that is permitted in a Type A
package, and A2 is the maximum
activity of normal form radioactive
material that is permitted in a Type A
package as prescribed in 10 CFR 71.4
and 49 CFR 173.403. The NRC’s and the
DOT’s transportation regulations
include package activity limits based on
fractions or multiples of the A1 and A2
values (e.g., 10¥3A2 and 3,000A2,
respectively).
In its concurrent harmonization
rulemaking, the DOT is proposing to
make changes to 49 CFR 173.435,
‘‘Table of A1 and A2 values for
radionuclides,’’ and 173.436, ‘‘Exempt
material activity concentrations and
exempt consignment activity limits for
radionuclides,’’ by adding seven
radionuclides, including barium-135m,
germanium-69, iridium-193m, nickel57, strontium-83, terbium-149, and
terbium-161. The NRC is proposing to
make corresponding changes to Tables
A–1 and A–2 to add these
radionuclides. The NRC is proposing to
revise the specific activity of natural
rubidium (Rb(nat)) to correct an error
that was introduced in the 1995 version
of the rule. Table A–1 of Appendix A to
10 CFR part 71 gives the specific
activity as 6.7 × 106 TBq/g, 1.8 × 108 Ci/
g. However, the correct value for the
specific activity of Rb(nat) is 670 Bq/g
(6.7 × 10¥10 TBq/g, 1.8 × 10¥8 Ci/g).
The A1 and A2 values were not
impacted by this error and remain
correct. The NRC is also proposing to
revise footnote c at the end of Table A–
2 to state that in the case of thoriumnatural, 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.
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Issue 15.4. Revision to Agreement State
Compatibility Categories
The NRC is proposing several changes
to the compatibility category
designations related to the QAP and
reporting requirements. These changes
would ensure that Agreement States
have the appropriate authority to
approve, inspect, and enforce QAPs for
their licensees, as well as that the NRC
and Agreement States receive important
reports regarding issues with radioactive
material shipments.
The NRC is proposing to revise the
compatibility category designations for
the regulations containing QAP
requirements for those Agreement States
that have licensees located within their
States who use NRC-approved Type B
packages, other than for industrial
radiography, to ship Type B quantities
of radioactive material; or have
licensees that ship using the general
license in § 71.21, ‘‘General license: Use
of foreign approved package’’; § 71.22,
‘‘General license: Fissile material’’; or
§ 71.23, ‘‘General license: Plutoniumberyllium special form material.’’ The
NRC is also proposing to revise the
compatibility category designation for
the reporting requirements in § 71.95.
In the 2004 final rule (69 FR 3697;
January 26, 2004) that revised § 71.101,
‘‘Quality assurance requirements,’’ the
NRC stated that § 71.101(b), and (c)(1)
are designated as Compatibility
Category C for those Agreement States
that have licensees that use Type B
packages, other than for industrial
radiography. For Compatibility Category
C, the essential objectives of the NRC
program elements should be adopted by
such Agreement States. The NRC is
proposing to change the compatibility
category designation for 71.101(b) and
(c)(1) from C to B. This is consistent
with Management Directive 5.9,
‘‘Adequacy and Compatibility of
Program Elements for Agreement State
Programs,’’ which states that program
elements in Compatibility Category B
are those that apply to activities that
cross jurisdictional boundaries. Since
the QAP activities in 71.101(b) and
(c)(1) are used during domestic shipping
of radioactive material and therefore
cross jurisdictional boundaries, a B
compatibility would align with
Management Directive 5.9 criteria. Also,
many of the regulations that contain
QAP review criteria (e.g., §§ 71.109,
71.111, 71.113, 71.115, 71.117, 71.119,
71.121, 71.123, and 71.125) were
addressed in the 2004 rule, but were
designated as Compatibility Category
NRC, which relate to areas of regulation
reserved to the NRC that cannot be
adopted by the Agreement States. The
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NRC is proposing to address these
compatibility issues in this proposed
rule so that, consistent with the intent
of the 2004 rulemaking, Agreement
States can adopt compatible QAP
regulations that would require their
licensees to follow these QAP criteria
and allow Agreement States to approve,
inspect and enforce their licensees’
QAPs. Specifically, this rule proposes to
correct the compatibility category
designation to B for many of these
regulations that are currently
Compatibility Category NRC, C, or D.
This change would require Agreement
States to have essentially identical
regulations and would give the
Agreement States the authority to
approve, inspect and enforce their
licensees’ QAPs. Only Agreement States
with licensees that use Type B packages,
other than for industrial radiography, or
with licensees that ship using the
general license in § 71.21, § 71.22, or
§ 71.23, which also requires an
approved QAP, would be impacted.
Additionally, the regulations in
§ 71.95 require NRC licensees to submit
a written report to the NRC of instances
in which there is a significant reduction
in the effectiveness of any NRCapproved package; details of defects
with safety significance in any NRCapproved package, after first use; and
instances in which the conditions of a
CoC were not followed during
shipment. In the 2004 final rule (69 FR
3697; January 26, 2004) that revised
§ 71.95, the NRC stated that the
compatibility category for § 71.95 is
Category D; therefore, it does not need
to be adopted by the Agreement States
to be compatible with the NRC’s
regulatory program. The reporting
requirements in § 71.95(a) are to ensure
that the NRC is alerted to instances in
which a package may have a defect or
has a significant reduction in
effectiveness such that, as needed, other
licensees authorized to use the package
are made aware of the possible issues.
Agreement State licensees also use NRCapproved packages, including industrial
radiography devices, but are not subject
to any of the requirements in § 71.95
and, therefore, are not required to
submit a report to the NRC pursuant to
§ 71.95. The NRC is proposing to change
the compatibility category for § 71.95(a)
to Compatibility Category C in order to
have Agreement State regulations
require notification to the NRC of these
instances. This will clarify that if a State
licensee uses an NRC-approved package
that has a defect or has a significant
reduction in effectiveness the NRC is
aware such that others using the
package can be made aware of the
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situation. The NRC also is proposing to
update the compatibility category for
§ 71.95(b) to Compatibility Category C to
ensure that the Agreement State agency
receives these reports from its licensees
indicating instances when the CoC was
not followed. As noted in the 1995 final
rule (60 FR 50248, 50259), the purpose
of this requirement is to provide
feedback on QAP effectiveness.
Consistent with the compatibility
category corrections for other QAP
related regulations, this proposed rule
would also correct the compatibility
category for § 71.95(b) so that
Agreement States receive these QAPrelated reports. The compatibility
categories for § 71.95(c) and (d) would
also be revised to Compatibility
Category C so that these reports contain
the required information.
In summary, the NRC is proposing to
revise the compatibility category for (1)
§ 71.101(b) and (c)(1) from a
Compatibility Category C to B to be in
alignment with the criteria in
Management Directive 5.9; (2) many of
the QAP-related regulations (e.g.,
§§ 71.109, 71.111, 71.113, 71.115,
71.117, 71.119, 71.121, 71.123, and
71.125) from a Compatibility Category
NRC, C, or D to a B to allow the
Agreement States the authority to
approve, inspect and enforce these
regulations; and (3) the reporting
requirements in § 71.95(a) and (b) from
a Compatibility Category D to C so that
the NRC receives reports from
Agreement State licensees on package
defects pursuant to § 71.95(a), and that
Agreement State regulators receive
reports when their licensees do not use
an NRC-approved package in
accordance with the CoC pursuant to
§ 71.95(b), and to § 71.95(c) and (d) so
that these reports contain the required
information.
Issue 15.5. Deletion of Redundant
Advance Notification Requirements for
Shipment of Spent Nuclear Fuel
Section 71.97 is titled ‘‘Advance
notification of shipment of irradiated
reactor fuel and nuclear waste.’’
However, advance notification
requirements for irradiated reactor fuel
(and, equivalently, spent nuclear fuel)
are separately included in the more
general requirements of 10 CFR part 73,
‘‘Physical protection of plants and
materials.’’ Specifically, as required in
§ 73.37(b)(2), licensees are required to
provide advance notification of
shipment to the Governor of a State and/
or Tribal official for any shipment
crossing the State or Tribal boundary
when the shipment contains greater
than 100 grams irradiated reactor fuel
and the external radiation dose rate is
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greater than 1 Gy (100 rad) per hour at
a distance of 1 meter (3.3 feet) from any
accessible surface without intervening
shielding. Licensees are also required to
provide notification of such shipments
to the NRC in accordance with § 73.72.
Additionally, as required in § 73.35,
‘‘Requirements for physical protection
of irradiated reactor fuel (100 grams or
less) in transit,’’ licensees who transport
100 grams or less of irradiated reactor
fuel, when the external radiation dose
rate is greater than 1 Gy (100 rad) per
hour at a distance of 1 meter (3.3 feet)
from any accessible surface without
intervening shielding, are required to
provide advance notification of
shipment in accordance with § 37.77.
When 10 CFR part 37 was established in
2013, this requirement was introduced,
but the ‘‘irradiated reactor fuel’’ aspect
was not removed from § 71.97.
Therefore, licensees may need to
produce two reports for a single
shipment to meet the advance
notification requirements of §§ 71.97
and 73.37 or § 73.35. To address this
potential inefficiency the NRC is
proposing to modify § 71.97 to remove
references to irradiated reactor fuel.
IV. Specific Request for Comment
The NRC is seeking comment and
feedback from the public on this
proposed rule. The NRC is particularly
interested in comment and supporting
rationale from the public on the
following:
QUESTION 1: IAEA Changes in SSR–6
(2018 Edition) Not in the Scope of This
Proposed Rule
Starting in 2016, while developing the
regulatory basis for this proposed rule,
the NRC considered the changes in
SSR–6, 2012 Edition, and the proposed
changes that were being considered for
SSR–6, 2018 Edition, which were
eventually issued in June 2018. The
NRC contracted with Oak Ridge
National Laboratory (ORNL) to develop
ORNL/TM–2014/658, ‘‘Comparison of
the International and United States
Domestic Radioactive Material
Transport Regulations.’’ In this
document, ORNL compared both NRC
and DOT regulations to SSR–6, 2012
Edition, and noted the differences. The
NRC then compared the changes
between SSR–6, 2018 Edition, and the
2012 Edition to determine which
changes affect NRC regulations and
whether those changes should be
included in this proposed rule. Based
on this review, the NRC did not include
the following IAEA changes in the scope
of this proposed rule:
1. Issue 1 consisted of four different
sub-issues: Issue No. 1a: New Fissile
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Exceptions in IAEA SSR–6, paragraph
417; Issue No. 1b: Competent AuthorityApproved Fissile Exception, SSR–6,
paragraph 417(f); Issue No. 1c: CSIControlled Fissile Material Packages,
SSR–6, paragraph 674; and Issue No. 1d:
Plutonium Shipments in Type A
Packages, SSR–6, paragraph 675.
For issue 1a, the NRC considered
whether to adopt the fissile exceptions
in paragraphs 417(c), without
consignment limits in paragraph 570(c);
the consignment limit in paragraph
570(d) associated with the package mass
limit in paragraph 417(d); and the
exception in paragraph 417(e) and its
associated exclusive use restriction in
paragraph 570(e), but with a mass limit
of 140 g instead of the IAEA mass limit
of 45 grams of fissile material from SSR–
6, 2018 Edition, into the NRC
regulations. The NRC chose not to adopt
the consignment limits in 570(c) and (d)
for the fissile exceptions in 417(c) and
417(d), respectively because
consignment limits do not prevent the
accumulation of packages on a transport
conveyance, as there is no limit to the
number of consignments that may be
present on a single conveyance.
Additionally, the accumulation on a
single conveyance of the number of
these packages required to approach
criticality is not credible.
After evaluation of Issue 1b, the NRC
is not proposing to add the new
‘‘competent authority-approved’’ fissile
exception in paragraph 417(f) into the
NRC regulations. If an NRC licensee
wished to ship a material that did not
meet the fissile material exemption or
general license criteria in 10 CFR part
71, and for which demonstration of
subcriticality in a package per the
requirements of §§ 71.55 and 71.59 is
deemed too burdensome, the licensee
could request a specific exemption
under § 71.12. The NRC notes that if an
NRC licensee submitted a ‘‘competent
authority-approved’’ exception, the
approval would include both NRC and
DOT reviews and issuance of the
exception and the NRC review and
findings would be similar to those of
either an exemption or 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
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limits, and a formula for determination
of a lower CSI, for packages which
withstand normal conditions of
transport while maintaining a larger
minimum external dimension of 30
centimeters. The IAEA SSR–6,
paragraph 674(c), contains the same CSI
calculation as paragraph 674(b), for
packages that withstand normal
conditions of transport while
maintaining a minimum external
dimension of 10 centimeters, with a
limit of 15 grams fissile material per
package.
The NRC does not propose to adopt
the changes in IAEA SSR–6, paragraph
674, because the NRC has determined
that the mass limits and other
requirements in §§ 71.22 and 71.23 are
appropriate for providing criticality
safety equivalent to packages approved
under the criticality safety requirements
of §§ 71.55 and 71.59. Adopting the
provisions of IAEA SSR–6 would result
in more restrictive mass limits for the
fissile material general licenses
authorized under 10 CFR part 71.
The NRC evaluated issue 1d, SSR–6,
paragraph 675, to add NRC
requirements for shipment of plutonium
in a nonfissile package, with
accumulation control provided by the
calculation of a CSI. This provision was
included in SSR–6, 2012 Edition but
without accumulation control. The
NRC’s fissile exemption in § 71.15(f) is
similar in that it limits the package to
1000 g of plutonium, of which not more
than 20 percent by mass may be
plutonium-239, plutonium-241, or any
combination of the two; however, the
NRC regulation does not include
accumulation control via a CSI
calculation. The NRC has determined
that the fissile exemption in § 71.15(f) is
safe without accumulation control, and
that there is no safety benefit to limiting
accumulation through the use of a CSI,
in order to be consistent with the IAEA
standards. Therefore, the NRC is not
proposing to harmonize with paragraph
675, SSR–6, 2018 Edition.
2. The NRC considered adopting the
reduced external pressure value of 60
kPa from paragraph 645 and the air
transport package requirements from
paragraph 621. The NRC is not
proposing to harmonize with paragraphs
621 and 645, SSR–6, 2018 Edition, as
discussed for Issue 2 in Section III of
this proposed rule, to avoid creating
unnecessary mode-specific restrictions
within 10 CFR part 71.
3. Inclusion of Type C Package
Standards (paragraphs 669–672)—The
NRC considered adding Type C package
standards for domestic transport, but
there was not an expressed need for
domestic transport of packages
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approved to Type C standards.
Therefore, the NRC is not proposing to
add Type C package standards in this
proposed rule.
4. Testing and reporting the integrity
of the containment system and
shielding, and assessing criticality
safety (paragraph 716), and additional
description of the impact of the tests on
packages (paragraphs 718–737)—The
NRC reviewed its regulations for an
application for approval of a package
design and considered its regulations
sufficient to obtain the information
needed to determine whether a package
design meets the requirements in 10
CFR part 71.
5. Addition of LSA Fissile Shipments
(paragraphs 518, 519, 520)—Since LSA
packages are self-certified under DOT
regulations, other than the fissile
material exemptions (§ 71.15) and fissile
material general licenses (§§ 71.22 and
71.23), there is no mechanism for
adding fissile material to an LSA
package without NRC approval. Under
current NRC regulations, the package
could be certified but would become a
Type BF or Type AF package,
depending on the quantity of
radioactive material in the package, and
therefore the NRC did not consider any
revision necessary.
6. Safety Factors for Lifting
Attachments (paragraph 608)—The NRC
regulations in § 71.45 contain
quantitative criteria for evaluating
lifting attachments that are considered a
structural part of the package. The IAEA
standards state an ‘‘appropriate’’ safety
factor must be used. In its review, the
NRC determined that adopting the IAEA
changes would not result in safety
benefits beyond those in § 71.45.
7. Shipment after Storage and Gap
Analysis (paragraphs 503(e) and
809(k))—The IAEA added regulations
both for shipment after storage and a
gap analysis for packages in storage
prior to shipment. The regulations in
SSR–6, paragraph 503(e), require that
during storage, packages are maintained
to ensure that all relevant transportation
standards in SSR–6 and certificates of
approval for those packages will be
fulfilled. The NRC is not proposing to
adopt paragraph 503(e) because, during
its review of packages for which storage
is expected prior to transport (i.e., dual
purpose casks or canisters), the NRC
ensures that the evaluations, operating
procedures, maintenance program and
acceptance tests for transport take
storage into consideration. In addition,
for any package that is stored prior to
transport, existing NRC requirements
(§§ 71.17(c) and 71.87(b)) ensure that,
prior to transport, the licensee must
comply with the terms and conditions
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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’’
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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
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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 § 71.106 to achieve
NRC’s stated intent in the 2015 final
rule. Specifically, the NRC is proposing
to revise § 71.106(b) to clarify that a
biennial report must be submitted to the
NRC even if no changes are made to the
QAP during the reporting period. This
proposed requirement would benefit the
NRC’s regulatory oversight of QAP
approval holders. The NRC inspection
program for 10 CFR part 71 QAP
approval holders relies on having
current information about the QAP
available to the NRC, including the
reporting of no changes. The 24-month
reporting period aims to provide an
appropriate balance between the burden
placed on the QAP approval holders
and the need to ensure that the NRC has
current information, especially when
considering most QAP approval holders
subject to periodic inspection are
inspected every 5 years or on an asneeded basis. Another benefit is that the
revised QAP reporting requirements in
10 CFR part 71 would be consistent
with those in 10 CFR 50.54(a)(3) and
50.71(e)(2) for 10 CFR part 50 QAPs.
The benefits and costs of the proposed
requirement are described in the
regulatory analysis and the NRC
estimates that the cost of compliance is
very small. The NRC is interested in the
public’s feedback as to the benefits and
costs of requiring a no-change biennial
report.
• 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.
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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.
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Section 71.19 Previously Approved
Package
This proposed rule would revise
paragraph (a) to include existing CoCs
that have a ‘‘–96’’ in their package
identification number, redesignate
paragraphs (c) and (d) as paragraphs (d)
and (e), revise newly redesignated
paragraph (e) to include those CoCs that
have a suffix ‘‘–96’’ in their
identification numbers, and add new
paragraph (c), to add transitional
arrangements on existing CoCs that have
a ‘‘–96’’ in their package identification
number.
Section 71.22 General License: Fissile
Material
This proposed rule would revise
paragraph (a) to replace ‘‘subparts E and
F of this part’’ with ‘‘§§ 71.55 and
71.59’’ and to remove the limitation to
a Type A quantity of radioactive
material in a Type A package to allow
shipment of material under the general
licenses in §§ 71.22 and 71.23 in a Type
B package, paragraph (c) to remove
(c)(1) and redesignate paragraph (c)(2) as
new paragraph (c), paragraphs (e)(3)
through (5) to limit the 233U to less than
one percent of the mass of 235U, similar
to the provision limiting plutonium in
§ 71.22(e)(5)(ii), and add new
paragraphs (f) through (h) to ensure that
each licensee will comply with
§ 71.17(c) for shipments made using the
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respective general license and that any
Type B package used under the
respective general license approved by
the NRC before the effective date of the
final rule is subject to the transitional
arrangements in § 71.19.
Section 71.23 General License:
Plutonium-Beryllium Special Form
Material
This proposed rule would revise
paragraphs (a) and (c), and add
paragraphs (f) through (h) to clarify that
only special form sealed sources, not
just sealed sources may be delivered to
a carrier for transport using the general
license in § 71.23.
Section 71.31
Contents of Application
This proposed rule would revise
paragraph (a) to add a maintenance
program description, as required by
§ 71.35 among the contents of
application.
Section 71.35
Package Evaluation
This proposed rule would revise
paragraph (b) to delete ‘‘and’’ paragraph
(c) to add ‘‘; and’’ and add new
paragraph (d) to specify maintenance
program requirements.
Section 71.43
All Packages
General Standards for
This proposed rule would revise
paragraph (d) to specifically include the
evaluation of the effects of aging, and to
specify that degradation evaluations
will be managed by the maintenance
program in accordance with § 71.35(d),
and add new paragraph (i) to specify
that each system designed to contain
liquids has adequate ullage during
evaluation of the tests and conditions
for normal conditions of transport and
hypothetical accident conditions
specified in §§ 71.71 and 71.73.
Section 71.55 General Requirements
for Fissile Material Packages
This proposed rule would revise
paragraph (g)(1) to require that there is
no contact between the cylinder plug
and any other part of the packaging,
other than at its original attachment
point and that the cylinder plug remains
leak tight, as NRC requires for the
cylinder valve.
Section 71.71
Transport
Normal Conditions of
This proposed rule would change the
unit of measure in the table in
paragraph (c)(1) to change the unit of
measure for the values of insolation
used for the heat test for normal
conditions of transport from ‘‘(g cal/
cm2)’’ to ‘‘(W/m2)’’.
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Section 71.73
Conditions
55723
Hypothetical Accident
This proposed rule would revise
paragraph (b) to add insolation to the
initial conditions for the tests for
hypothetical accident conditions.
Section 71.77
Material
Qualification of LSA—III
This proposed rule would remove and
reserve § 71.77 and make conforming
changes to §§ 71.4 and 71.100.
Section 71.95
Reports
This proposed rule would remove
paragraph (a)(3) as it is duplicative to
text in paragraph (b).
Section 71.97 Advance Notification of
Shipment of Irradiated Reactor Fuel and
Nuclear Waste
This proposed rule would revise the
section title, the introductory text of
paragraph (b), and paragraphs (d) and
(f)(1) to remove references to irradiated
reactor fuel to correct a duplicative
advance notification reporting
requirement in § 71.97 with those in
§§ 73.35 and 73.37.
Section 71.100
Criminal Penalties
This proposed rule would revise
paragraph (b) to remove the leaching
test requirement as a conforming change
to § 71.77.
Section 71.106 Changes to Quality
Assurance Program
This proposed rule would revise the
introductory text of paragraph (b) to
clarify that a biennial report must be
submitted to the NRC even if no changes
are made to the QAP during the
reporting period.
Appendix A to Part 71—Determination
of A1 and A2
This proposed rule would revise
Tables A–1 and A–2 in paragraph V.b.
to add seven radionuclides and correct
the specific activity of natural rubidium.
VI. Regulatory Flexibility Certification
Under the Regulatory Flexibility Act
(5 U.S.C. 605(b)), the NRC certifies that
this proposed rule will not, if issued,
have a significant economic impact on
a substantial number of small entities.
This proposed rule affects a number of
‘‘small entities’’ as defined by the
Regulatory Flexibility Act or the size
standards established by the NRC
(§ 2.810). However, as indicated in the
regulatory analysis, these amendments
do not have a significant economic
impact on the affected small entities.
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Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / Proposed Rules
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.
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VIII. Backfitting and Issue Finality
The NRC has determined that
backfitting (§ 50.109, § 70.76, § 72.62, or
§ 76.76) and the issue finality provisions
in 10 CFR part 52 do not apply to this
proposed rule because it would not
involve any provisions that would
impose backfits as defined in 10 CFR
chapter I or affect the issue finality of
any approval issued under 10 CFR part
52. Some licensees that are within the
scope of the backfit rule (e.g., a power
reactor or a fuel fabrication facility)
transport radioactive material from their
own facilities. Those backfitting and
issue finality provisions apply to
activities directly regulated under those
parts, and do not apply to activities
regulated under other parts that do not
include backfitting or issue finality
provisions. The exception to this
general principle is where the activity
regulated under other parts that do not
include backfitting or issue finality
provisions is an inextricable part of the
regulated activity within the scope of
backfitting or issue finality. Preparing
packages for transport is not an
inextricable part of the procedures or
organization required to design,
construct or operate a facility as
licensed under 10 CFR part 50, 52, 70,
72, or 76; rather, it is a separate activity
that these licensees may choose to
undertake. The scope of this proposed
rule does not include any changes to
any of those facilities or plants’
activities for which the backfit rule
applies.
The NRC’s determination on this
matter is in accordance with
Management Directive 8.4,
‘‘Management of Backfitting, Forward
Fitting, Issue Finality, and Information
Requests,’’ and its associated guidance
in NUREG–1409, ‘‘Backfitting
Guidelines.’’
IX. Cumulative Effects of Regulation
The NRC seeks to minimize any
potential negative consequences
resulting from the cumulative effects of
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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
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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
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Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / Proposed Rules
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
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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
Infocollects.Resource@nrc.gov.
• 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 § 71.100. One of the actions
within the scope of this rulemaking,
Issue 6, Deletion of the Low Specific
Activity—III Leaching Test, proposes to
remove the content of § 71.77 and
replace the section heading with
‘‘RESERVED.’’ This change would
impact § 71.100(b), because § 71.77
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55725
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
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Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / Proposed Rules
Agreement States under the Atomic
Energy Act of 1954, as amended, or
provisions of title 10 of the Code of
Federal Regulations. These program
elements should not be adopted by the
Agreement States. Adequacy category
H&S program elements are program
elements that are required because of a
particular health and safety role in the
regulation of agreement material within
the State and should be adopted in a
manner that embodies the essential
objectives of the NRC program. A
bracketed compatibility category (e.g.,
[B]) means that the provision may have
been adopted elsewhere in the
Agreement State’s regulations and does
not need to be adopted again.
As discussed in Section III of this
document, Issue 15.4, the regulations
that contain QAP requirements (e.g.,
§§ 71.109, 71.111, 71.113, 71.115,
71.117, 71.119, 71.121, 71.123, and
71.125) are currently designated as
Compatibility Category NRC and cannot
be adopted by the Agreement States.
Since a proper QAP review cannot be
completed without addressing many of
these criteria, Agreement States would
need to adopt compatible regulations to
require licensees that use NRC-approved
Type B packages for shipping, other
than for industrial radiography, or that
ship using the general license in § 71.21,
§ 71.22 or § 71.23, to follow these QAP
criteria. Additionally, since only a few
Agreement States have applicable
licensees that perform shipments of
Type B quantities of radioactive
materials, other than for industrial
radiography operations (which are
covered under § 34.31), or that ship
using the general license in § 71.21,
§ 71.22, or § 71.23, all QAP-related
requirements, including those
mentioned previously and others
referenced below in the table, would be
re-designated as a Compatibility
Category B. This re-designation would
require those Agreement States with
applicable licensees to have essentially
identical regulations. For those
Agreement States that do not have
applicable licensees, these regulations
will remain designated as Compatibility
Category D and, hence, do not have to
be adopted for purposes of
compatibility.
The changes in this proposed rule,
discussed in Section III of this
document, would be a matter of
compatibility between the NRC and the
Agreement States, thereby providing
consistency among Agreement State and
NRC requirements. Regulations that are
a part of this rulemaking but remain the
same compatibility category designation
are included in the table for
completeness. The compatibility
categories are designated in the
following table.
Compatibility
Section
Change
Subject
Existing
71.0(d)(1) .............................................
71.4 ......................................................
71.4 ......................................................
Revised ...........................
New .................................
Revised ...........................
71.4 ......................................................
71.4 ......................................................
71.15(a) and (d) ..................................
71.15(g) ...............................................
71.17(e) ...............................................
71.19 ....................................................
71.22(a), (c), and (e)(3) through (5) ....
71.22(f) through (h) .............................
71.23(a) and (c) ...................................
Revised ...........................
Revised ...........................
Revised ...........................
New .................................
Revised ...........................
Revised ...........................
Revised ...........................
New .................................
Revised ...........................
71.23(f) through (h) .............................
New .................................
71.31(a) ...............................................
71.35(b) and (c) ...................................
71.35(d) ...............................................
71.43(d) ...............................................
71.43(i) ................................................
71.55(g) ...............................................
71.71(c)(1) ...........................................
71.73(b) ...............................................
71.77 ....................................................
71.95 ....................................................
Revised ...........................
Revised ...........................
New .................................
Revised ...........................
New .................................
Revised ...........................
Revised ...........................
Revised ...........................
Removed ........................
Revised compatibility category.
Removed ........................
Revised ...........................
71.95(a)(3) ...........................................
71.97 ....................................................
71.100 ..................................................
71.101(b) .............................................
71.101(c)(1) .........................................
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71.103(a) and (b) ................................
71.103(c), (d), (e) and (f) ....................
71.105 ..................................................
71.106 ..................................................
71.109 ..................................................
VerDate Sep<11>2014
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Revised ...........................
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
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Purpose and Scope .................................................
Definition: Radiation Level .......................................
Definition: Low Specific Activity (LSA) material [Deletion of Low Specific Activity—III Leaching Test].
Definition: Special form radioactive material ...........
Definition: Surface Contaminated Object (SCO) .....
Exemption from classification as fissile material .....
Exemption from classification as fissile material .....
General license: NRC-approved package ...............
Previously approved package ..................................
General license: Fissile material ..............................
General license: Fissile material ..............................
General license: Plutonium-beryllium special form
material.
General license: Plutonium-beryllium special form
material.
Contents of application ............................................
Package evaluation ..................................................
Package evaluation ..................................................
General standards for all packages .........................
General standards for all packages .........................
General requirements for fissile material packages
Normal conditions of transport .................................
Hypothetical accident conditions .............................
Qualification of LSA—III Material .............................
Reports .....................................................................
D
[B]
[B]
[B]
[B]
B
NRC
[B]
[B]
New
D
[A]
[B]
[B]
[B]
[B]
[B]
B
NRC
[B]
[B]
[B]
[B]
NRC
NRC
NRC
NRC
NRC
NRC
NRC
D
NRC
NRC
NRC
NRC
NRC
NRC
NRC
NRC
** C
Reports .....................................................................
Advance notification of shipment of irradiated reactor fuel and nuclear waste.
Criminal penalties ....................................................
Quality assurance requirements ..............................
D
B
*
B
D
*** C
D
*** B
Quality assurance requirements ..............................
*** C
** B
Quality assurance organization ...............................
*** C
** B
Quality assurance organization ...............................
D
** B
Quality assurance program ......................................
C
** B
Changes to quality assurance program ...................
C
** B
Procurement document control ................................
NRC
** B
Fmt 4702
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Compatibility
Section
Change
Subject
Existing
71.111 ..................................................
71.113 ..................................................
71.115 ..................................................
71.117 ..................................................
71.119 ..................................................
71.121 ..................................................
71.123 ..................................................
71.125 ..................................................
71.127 ..................................................
71.129 ..................................................
71.131 ..................................................
71.133 ..................................................
71.135 ..................................................
71.137 ..................................................
Table A–1 in Appendix A to 10 CFR
Part 71.
Table A–2 in Appendix A to 10 CFR
Part 71.
New
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised compatibility category.
Revised ...........................
Instructions, procedures and drawings ....................
NRC
** B
Document control .....................................................
NRC
** B
Control of purchased material, equipment, and
services.
Identification and control of materials, parts and
components.
Control of special processes ...................................
NRC
** B
NRC
** B
NRC
** B
Internal inspection ....................................................
NRC
** B
Test control ..............................................................
NRC
** B
Control of measuring and test equipment ...............
NRC
** B
Handling, storage, and shipping control ..................
[C]
** B
Inspection, test, and operating status ......................
[C]
** B
Nonconforming materials, parts, or components .....
[C]
** B
Corrective action ......................................................
C
** B
Quality assurance records .......................................
*** C
** C
Audits .......................................................................
C
** C
A1 and A2 Values for Radionuclides ........................
[B]
[B]
Revised ...........................
Exempt Material Activity Concentrations and Exempt Consignment Activity Limits for Radionuclides.
[B]
[B]
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* Denotes regulations that are designated Compatibility Category D but which will be removed from the regulations as a result of these proposed amendments. Agreement States that have an equivalent regulation should remove these provisions from their regulations when the regulations become final.
** B/C (as designated)—for Agreement States that have licensees that use Type B approved packages for shipping, other than for industrial radiography, or have licensees that ship using the general license in § 71.21, § 71.22, or § 71.23, these regulations are required for compatibility
purposes.
D—for States that do not have licensees that use Type B approved packages for shipping, other than for industrial radiography, these regulations are not required for compatibility purposes.
*** 10 CFR 71.101(g) indicates that QA programs for industrial radiography Type B package users are covered by § 34.31(b). It also indicated
that this section satisfies § 71.17(b) and therefore will satisfy those sections referenced in this provision (§§ 71.101 through 71.137).
The NRC invites comment on the
compatibility category designations in
the proposed rule and suggests that
commenters refer to Handbook 5.9 of
Management Directive 5.9, ‘‘Adequacy
and Compatibility of Program Elements
for Agreement State Programs,’’ for more
information. The NRC notes that, like
the rule text, the compatibility category
designations can change between the
proposed rule and final rule on the basis
of comments received and Commission
decisions regarding the final rule. The
NRC encourages anyone interested in
commenting on the compatibility
category designations to do so during
the comment period.
XVI. Voluntary Consensus Standards
The National Technology Transfer
and Advancement Act (NTTAA) of
1995, Public Law 104–113, requires that
Federal agencies use technical standards
that are developed or adopted by
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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
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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
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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/publicmeetings/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 link/
Federal Register citation
Document
lotter on DSK11XQN23PROD with PROPOSALS1
Rulemaking Documents and References
SECY–20–0102 for this proposed rule ...........................................................................................
Federal Register notice for this proposed rule ..............................................................................
Regulatory Analysis for this proposed rule .....................................................................................
Environmental Assessment for this proposed rule .........................................................................
OMB supporting statement for this proposed rule ..........................................................................
Draft regulatory basis document for this rulemaking, dated March 2019 ......................................
Federal Register notification for draft regulatory basis, dated April 12, 2019 ..............................
Draft regulatory basis comment submission #1 ..............................................................................
Draft regulatory basis comment submission #2 ..............................................................................
Draft regulatory basis comment submission #3 ..............................................................................
Draft regulatory basis comment submission #4 ..............................................................................
Draft regulatory basis comment submission #5 ..............................................................................
Draft regulatory basis comment submission #6 ..............................................................................
Draft regulatory basis comment submission #7 ..............................................................................
NRC final rule amending packaging and transportation of radioactive material regulations,
dated June 12, 2015.
DOT final rule amending packaging and transportation of radioactive material regulations,
dated July 11, 2014.
NRC final rule harmonizing its regulations with the 1996 edition of IAEA Safety Series No. 6,
dated January 26, 2004.
NRC proposed rule harmonizing its regulations with the 1996 edition of IAEA Safety Series No.
6, dated April 30, 2002.
NRC final rule harmonizing its regulations with the 1985 edition of IAEA Safety Series No. 6,
dated September 28, 1995.
NRC/DOT Memorandum of Understanding, dated July 2, 1979 ....................................................
SECY–16–0093, ‘‘Rulemaking Plan for Revisions to Transportation Safety Requirements and
Harmonization with International Atomic Energy Agency Transportation Requirements,’’ dated
July 28, 2016.
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ML19143A311
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ML19148A147
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79 FR 40589
69 FR 3697
67 FR 21390
60 FR 50248
44 FR 38690
ML16158A164
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ADAMS accession No./web link/
Federal Register citation
Document
Staff Requirements Memorandum SRM–SECY–16–0093, ‘‘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 Potential Revisions to Transportation Safety Requirements and Harmonization with International Atomic Energy Agency Transportation Requirements,’’ dated November 15, 2016.
Federal Register notification for harmonization issues paper, dated November 21, 2016 ..........
Issues paper public meeting summary, ‘‘Summary of the December 5 and 6, 2016 Public Meeting on Issues Paper on Revisions to Transportation Safety Requirements and Harmonization
with the International Atomic Energy Agency Transportation Requirements,’’ dated December
14, 2016.
ML16235A182
ML16299A298 paper, ML16299A291 package
81 FR 83171
ML16343A661
Draft Regulatory Guidance Document
Draft Regulatory Guide DG–7011, ‘‘Standard Format and Content of Part 71 Applications for
Approval of Packages for Radioactive Material,’’ Revision 3 of Regulatory Guide 7.9.
ML22223A085
IAEA Transportation Safety Standards and Related References
SSR–6, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 2018 Edition ...................
SSR–6, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 2012 Edition ...................
TS–R–1, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 2009 Edition ..................
TS–R–1, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 2005 Edition ..................
TS–R–1, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 1996 Edition ..................
Safety Series No. 6, ‘‘Regulations for the Safe Transport of Radioactive Material, 1985 Edition
(As Amended in 1990)’’.
Safety Series No. 6, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 1985 Edition
Safety Series No. 6, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 1973 Edition
Safety Series No. 6, ‘‘Regulations for the Safe Transport of Radioactive Material,’’ 1967 Edition
https://www.iaea.org/publications/12288/regulations-for-the-safe-transport-of-radioactive-material
https://www.iaea.org/publications/8851/regulations-for-the-safe-transport-of-radioactive-material-2012-edition
https://www.iaea.org/publications/8005/regulations-for-the-safe-transport-of-radioactive-material-2009-edition
https://www.iaea.org/publications/7291/regulations-for-the-safe-transport-of-radioactive-material-2005-edition
https://www.iaea.org/publications/6056/regulations-for-the-safe-transport-of-radioactive-material-1996-edition-revised
https://gnssn.iaea.org/Superseded%20
Safety%20Standards/Safety_Series_006_
1990.pdf
https://gnssn.iaea.org/Superseded%20
Safety%20Standards/Safety_Series_006_
1985.pdf
https://gnssn.iaea.org/Superseded%20
Safety%20Standards/Safety_Series_006_
1973.pdf
https://gnssn.iaea.org/Superseded%20
Safety%20Standards/Safety_Series_006_
1967.pdf
Other International Standards References
ANSI N14.1–2012, ‘‘Nuclear Materials—Uranium Hexafluoride—Packagings for Transport,’’
dated December 3, 2012.
ANSI N14.5–2014, ‘‘American National Standard for Radioactive Materials—Leakage Tests on
Packages for Shipment,’’ dated June 19, 2014.
International Organization for Standardization 7195:2005, ‘‘Nuclear Energy—Packaging of Uranium Hexafluoride (UF6) for Transport,’’ dated September 2005.
American National Standards Institute/American Nuclear Society 8.1–2014 (Reaffirmed 2018),
‘‘Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,’’ American Nuclear Society, La Grange Park, IL.
https://webstore.ansi.org/standards/pcc/
ansin142012
https://webstore.ansi.org/standards/pcc/
ansin142014
https://www.iso.org/standard/31251.html
https://webstore.ansi.org/Standards/ANSI/
ANSIANS2014R2018
Miscellaneous References
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National Renewable Energy Laboratory Solar Radiation Data ......................................................
NRC letter to Agreement States, ‘‘Clarification of Title 10 of the Code of Federal Regulations,
Part 71 Requirements Identified in Regulation Amendment Tracking System Identification
Number RATS ID: 2015–3 (STC–17–060),’’ dated August 15, 2017.
Presidential Memorandum, ‘‘Plain Language in Government Writing,’’ published June 10, 1998
Agreement State Program Policy Statement, dated October 18, 2017 .........................................
NRC Management Directive 5.9, Handbook 5.9, ‘‘Adequacy and Compatibility of Program Elements for Agreement State Programs,’’ dated April 26, 2018.
NRC Management Directive 8.4, ‘‘Management of Backfitting, Forward Fitting, Issue Finality,
and Information Requests,’’ dated September 20, 2019.
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82 FR 48535
ML18081A070
ML18093B087
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Document
ADAMS accession No./web link/
Federal Register citation
ORNL/TM–2014/658, ‘‘Comparison of the International and United States Domestic Radioactive
Material Transport Regulations,’’ dated September 30, 2014.
NUREG–1409, ‘‘Backfitting Guidelines,’’ Revision 1, draft for public comment, dated March
2020.
NUREG–1608, ‘‘Categorizing and Transporting Low Specific Activity Materials and Surface
Contaminated Objects,’’ dated July 1998.
NUREG–1660, ‘‘U.S.-Specific Schedules of Requirements for Transport of Specified Types of
Radioactive Material Consignments,’’ dated January 1999.
NUREG–1886, ‘‘Joint Canada–United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages,’’ dated March 2009.
NUREG–2216, ‘‘Standard Review Plan for Transportation Packages for Spent Fuel and Radioactive Material,’’ dated August 2020.
https://rampac.energy.gov/docs/default-source/
doeinfo/ORNL-TM-2014-658.pdf
ML18109A498
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
1. The authority citation for part 71
continues to read as follows:
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■
Authority: Atomic Energy Act of 1954,
secs. 53, 57, 62, 63, 81, 161, 182, 183, 223,
234, 1701 (42 U.S.C. 2073, 2077, 2092, 2093,
2111, 2201, 2232, 2233, 2273, 2282, 2297f);
Energy Reorganization Act of 1974, secs. 201,
202, 206, 211 (42 U.S.C. 5841, 5842, 5846,
5851); Nuclear Waste Policy Act of 1982, sec.
180 (42 U.S.C. 10175); 44 U.S.C. 3504 note.
Section 71.97 also issued under Sec. 301,
Pub. L. 96–295, 94 Stat. 789 (42 U.S.C. 5841
note).
2. In § 71.0, revise paragraph (d)(1) to
read as follows:
■
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§ 71.0
Purpose and scope.
*
*
*
*
*
(d)(1) Exemptions from the
requirement for license in § 71.3 are
specified in § 71.14. The general license
in § 71.21 does not require NRC package
approval. The general licenses in
§§ 71.22 and 71.23 require NRC package
approval if the quantities exceed a Type
A quantity. The general license in
§ 71.17 requires that an NRC certificate
of compliance or other package approval
be issued for the package to be used
under this general license.
*
*
*
*
*
■ 3. Amend § 71.4 by:
■ a. Revising the definitions for Low
Specific Activity material and Special
form radioactive material;
■ b. Revising the introductory text and
add paragraph (3) for Surface
contaminated object; and
■ c. Adding the definition Radiation
level in alphabetical order.
The revisions and addition read as
follows:
§ 71.4
Definitions.
*
*
*
*
*
Low Specific Activity (LSA) material
means radioactive material with limited
specific activity which is nonfissile or is
exempt under § 71.15, and which
satisfies the descriptions and limits set
forth in the following section. Shielding
materials surrounding the LSA material
may not be considered in determining
the estimated average specific activity of
the package contents. The LSA material
must be in one of three groups:
*
*
*
*
*
(3) LSA—III. Solids (e.g., consolidated
wastes, activated materials), excluding
powders, in which:
(i) The radioactive material is
distributed throughout a solid or a
collection of solid objects, or is
essentially uniformly distributed in a
solid compact binding agent (such as
concrete, bitumen, ceramic, etc.); and
(ii) [Reserved]
(iii) The estimated average specific
activity of the solid, excluding any
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shielding material, does not exceed 2 ×
10¥3A2/g.
*
*
*
*
*
Radiation level means the radiation
dose equivalent rate expressed in
millisieverts per hour or mSv/h
(millirems per hour or mrem/h).
*
*
*
*
*
Special form radioactive material
means radioactive material that satisfies
the following conditions:
(1) It is either a single solid piece or
is contained in a sealed capsule that can
be opened only by destroying the
capsule;
(2) The piece or capsule has at least
one dimension not less than 5 mm (0.2
in); and
(3) It satisfies the requirements of
§ 71.75. A special form encapsulation
designed in accordance with the
requirements of § 71.4 in effect from
April 1, 1996, to September 30, 2004,
may continue to be used, provided that
fabrication of the special form
encapsulation was successfully
completed by [DATE ONE DAY PRIOR
TO EFFECTIVE DATE OF FINAL
RULE]. A special form encapsulation
designed in accordance with the
requirements of § 71.4 in effect from
October 1, 2004, to [DATE ONE DAY
PRIOR TO EFFECTIVE DATE OF FINAL
RULE] may continue to be used,
provided that fabrication of the special
form encapsulation is successfully
completed by December 31, 2025. Any
other special form encapsulation must
meet the specifications of this
definition.
*
*
*
*
*
Surface contaminated object (SCO)
means a solid object that is not itself
classed as radioactive material, but
which has radioactive material
distributed on any of its surfaces. SCO
must be in one of three groups with
surface activity not exceeding the
following limits:
*
*
*
*
*
(3) SCO—III: A large solid object
which, because of its size, cannot be
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transported in a type of package
described in 49 CFR 173.403 of the DOT
regulations and for which:
(i) All openings are sealed to prevent
release of radioactive material during
conditions defined in 49 CFR
173.427(d);
(ii) The inside of the object is as dry
as practicable;
(iii) The nonfixed contamination on
the external surface does not exceed the
contamination limits specified in the
DOT regulations in 49 CFR 173.443; and
(iv) The nonfixed contamination plus
the fixed contamination on the
inaccessible surface averaged over 300
cm2 does not exceed 8 × 105 Bq/cm2 (20
microcuries/cm2) for beta and gamma
emitters and low toxicity alpha emitters,
or 8 × 104 Bq/cm2 (2 microcuries/cm2)
for all other alpha emitters.
*
*
*
*
*
■ 4. In § 71.15, revise the introductory
text and paragraphs (a) and (d) and add
paragraph (g) to read as follows:
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§ 71.15 Exemption from classification as
fissile material.
Fissile material meeting the
requirements of at least one of the
paragraphs (a) through (g) of this section
are exempt from classification as fissile
material and from the fissile material
package standards of §§ 71.55 and 71.59
but are subject to all other requirements
of this part, except as noted.
(a) Individual package containing:
(1) 2 grams or less fissile material, or
(2) 3.5 grams or less uranium-235,
provided the uranium is enriched in
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 uranium233 content not exceeding 1 percent of
the mass of uranium-235, provided that
the mass of any beryllium, graphite, and
hydrogenous material enriched in
deuterium constitutes less than 5
percent of the uranium mass, and that
the fissile material is distributed
homogeneously and does not form a
lattice arrangement within the package.
*
*
*
*
*
(g) Packages transported under
exclusive use on a conveyance
containing a total of 140 grams or less
fissile material.
■ 5. In § 71.17, revise paragraph (e) to
read as follows:
§ 71.17 General license: NRC-approved
package.
*
*
*
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*
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(e) For a Type B or fissile material
package, the design of which was
approved by NRC before [EFFECTIVE
DATE OF FINAL RULE], the general
license is subject to the additional
restrictions of § 71.19.
■ 6. Amend § 71.19 by:
■ a. Revising paragraph (a);
■ b. Redesignating paragraphs (c) and
(d) as paragraphs (d) and (e);
■ c. Adding new paragraph (c); and
■ d. Revising newly redesignated
paragraph (e).
The revisions and addition read as
follows:
§ 71.19
Previously approved package.
(a) A Type B(U) package, a Type B(M)
package, or a fissile material package,
previously approved by the NRC but
without the designation ‘‘–85’’ or ‘‘–96’’
in the identification number of the NRC
CoC, may be used under the general
license of § 71.17 with the following
additional conditions:
(1) Fabrication of the package is
satisfactorily completed by April 1,
1999, as demonstrated by application of
its model number in accordance with
§ 71.85(c);
(2) A serial number which uniquely
identifies each packaging which
conforms to the approved design is
assigned to and legibly and durably
marked on the outside of each
packaging; and
(3) Paragraph (a) of this section
expires [DATE 8 YEARS AFTER
EFFECTIVE DATE OF THE FINAL
RULE].
*
*
*
*
*
(c) A Type B(U) package, a Type B(M)
package, or a fissile material package
previously approved by the NRC with
the designation ‘‘–96’’ in the
identification number of the NRC CoC,
may be used under the general license
of § 71.17 with the following additional
conditions:
(1) Fabrication of the package must be
satisfactorily completed by January 1,
2029, as demonstrated by application of
its model number in accordance with
§ 71.85(c); and
(2) A package used for a shipment to
a location outside the United States,
after December 31, 2025, is subject to
multilateral approval, as defined in the
DOT’s regulations at 49 CFR 173.403.
*
*
*
*
*
(e) NRC will revise the package
identification number to designate
previously approved package designs
that were designated as AF, B(U), B(M),
B(U)F, B(M)F, B(U)–85, B(U)F–85,
B(M)–85, B(M)F–85, AF–85, B(U)–96,
B(U)F–96, B(M)–96, B(M)F–96, or AF–
96 as appropriate, with the
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55731
identification number suffix AF, B(U),
B(M), B(U)F, B(M)F, after receipt of an
application demonstrating that the
design meets the requirements of this
part.
■ 7. In § 71.22, revise paragraphs (a), (c),
and (e)(3) through (5) and add
paragraphs (f) through (h) to read as
follows:
§ 71.22
General license: Fissile material.
(a) A general license is issued to any
licensee of the Commission to transport
fissile material, or to deliver fissile
material to a carrier for transport, if the
material is shipped in accordance with
this section. The fissile material need
not be contained in a package which
meets the standards of §§ 71.55 and
71.59. However, the material must be
contained in a Type A or Type B
package, consistent with the quantity of
radioactive material in the package.
*
*
*
*
*
(c) The general license applies only
when a package’s contents contain less
than 500 total grams of beryllium,
graphite, or hydrogenous material
enriched in deuterium.
*
*
*
*
*
(e) * * *
(3) The values of X, Y, and Z used in
the CSI equation must be taken from
Table 71–1 or 71–2, as appropriate
based on criteria from § 71.22(e)(4) and
(5).
(4) If Table 71–2 is used to obtain the
value of X, then:
(i) The total mass of plutonium and
uranium-233 must not exceed 1 percent
of the mass of uranium-235;
(ii) Values for the terms in the
equation for uranium-233 and
plutonium must be assumed to be zero;
and
(iii) The value of the uranium
enrichment must be known and be less
than the enrichment value used from
Table 71–2.
(5) Table 71–1 values for X, Y, and Z
must be used to determine the CSI if:
(i) The total mass of plutonium and
uranium-233 exceeds 1 percent of the
mass of uranium-235;
(ii) The uranium is of unknown
uranium-235 enrichment or greater than
24 weight percent enrichment; or
(iii) Substances having a moderating
effectiveness (i.e., an average hydrogen
density greater than H2O) (e.g., certain
hydrocarbon oils or plastics) are present
in any form, except as polyethylene
used for packing or wrapping. * * *
*
*
*
*
*
(f) Each licensee using the general
license under paragraph (a) of this
section to transport a Type B quantity of
licensed material must use a package for
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which a license, CoC, or other approval
has been issued by the NRC, and must
comply with the provisions in
§ 71.17(c).
(g) For shipment of a Type B quantity
of licensed material, this general license
applies only when the package approval
authorizes use of the package under the
general license in § 71.17 or this general
license.
(h) For a Type B package, the design
of which was approved by NRC before
[EFFECTIVE DATE OF FINAL RULE],
this general license is subject to the
additional restrictions of § 71.19.
■ 8. In § 71.23, revise paragraph (a) and
the introductory text of paragraph (c)
and add paragraphs (f) through (h) to
read as follows:
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Package evaluation.
*
(a) A general license is issued to any
licensee of the Commission to transport
fissile material in the form of
plutonium-beryllium (Pu-Be) special
form sources, or to deliver Pu-Be special
form sources to a carrier for transport,
if the material is shipped in accordance
with this section. This material need not
be contained in a package which meets
the standards of §§ 71.55 and 71.59.
However, the fissile material must be
contained in a Type A or Type B
package, consistent with the quantity of
radioactive material in the package.
*
*
*
*
*
(c) The general license applies only
when a package’s contents contain less
than 1000 grams of plutonium, provided
that plutonium-239, plutonium-241, or
any combination of these radionuclides,
constitutes less than 240 grams of the
total quantity of plutonium in the
package.
*
*
*
*
*
(f) Each licensee using the general
license under paragraph (a) of this
section to transport a Type B quantity of
licensed material must use a package for
which a license, CoC, or other approval
has been issued by the NRC, and must
comply with the provisions in
§ 71.17(c).
(g) For shipment of a Type B quantity
of licensed material, this general license
applies only when the package approval
authorizes use of the package under the
general license in § 71.17 or this general
license.
(h) For a Type B package, the design
of which was approved by NRC before
[EFFECTIVE DATE OF FINAL RULE],
this general license is subject to the
additional restrictions of § 71.19.
■ 9. In § 71.31, revise paragraph (a) to
read as follows:
16:33 Sep 09, 2022
Contents of application.
(a) An application for an approval
under this part must include, for each
proposed packaging design, the
following information:
(1) A package description as required
by § 71.33;
(2) A package evaluation as required
by § 71.35;
(3) A maintenance program
description, as required by § 71.35; and
(4) A quality assurance program
description, as required by § 71.37, or a
reference to a previously approved
quality assurance program.
*
*
*
*
*
■ 10. In § 71.35, revise paragraphs (b)
and (c) and add paragraph (d) to read as
follows:
§ 71.35
§ 71.23 General license: Plutoniumberyllium special form material.
VerDate Sep<11>2014
§ 71.31
*
*
*
*
(b) For a fissile material package, the
allowable number of packages that may
be transported in the same vehicle in
accordance with § 71.59;
(c) For a fissile material shipment, any
proposed special controls and
precautions for transport, loading,
unloading, and handling and any
proposed special controls in case of an
accident or delay; and
(d) A maintenance program to assure
that the packaging will perform as
intended throughout its time in service.
The maintenance program must include
periodic testing requirements,
inspections, and replacement criteria
and schedules for replacement and
repairs of components on an as-needed
basis.
■ 11. In § 71.43, revise paragraph (d)
and add paragraph (i) to read as follows:
§ 71.43 General standards for all
packages.
*
*
*
*
*
(d) A package must be made of
materials and construction that assure
that there will be no significant
chemical, galvanic, or other reaction
among the packaging components,
among package contents, or between the
packaging components and the package
contents, including possible reaction
resulting from inleakage of water, to the
maximum credible extent. The effects of
the aging mechanisms and the behavior
of materials under irradiation must be
evaluated on package components to
show that their performance is not
significantly degraded or that
degradation will be managed by the
maintenance program in accordance
with § 71.35(d).
*
*
*
*
*
(i) Each system designed for holding
liquids must be designed, constructed,
and prepared for shipment so that under
PO 00000
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Sfmt 4702
the tests specified in §§ 71.71 and 71.73,
there would be adequate space to
accommodate variations in temperature
of the liquid, dynamic effects, and
filling dynamics.
■ 12. In § 71.55, revise paragraph (g)(1)
to read as follows:
§ 71.55 General requirements for fissile
material packages.
*
*
*
*
*
(g) * * *
(1) Following the tests specified in
§ 71.73 (‘‘Hypothetical accident
conditions’’), there is no physical
contact between the valve body or the
plug and any other component of the
packaging, other than at its original
point of attachment, and the valve and
plug remain leak tight;
*
*
*
*
*
■ 13. In § 71.71, in the table in
paragraph (c)(1), revise the heading of
the second column to read as follows:
§ 71.71
*
Normal conditions of transport.
*
*
(c) * * *
(1) * * *
*
*
INSOLATION DATA
* * * ..............
*
*
Total insolation for a 12-hour
period (W/m2)
*
*
*
*
*
*
*
*
14. In § 71.73, revise paragraph (b) to
read as follows:
■
§ 71.73
Hypothetical accident conditions.
*
*
*
*
*
(b) Test conditions. Except for the
water immersion test, the following
conditions shall apply before and after
the tests:
(1) The ambient air temperature shall
remain constant at that value between
¥29 °C (¥20 °F) and +38 °C (+100 °F)
which is most unfavorable for the
feature under consideration;
(2) The insolation shall be that value
between 0 and the maximum value
listed in the Insolation Data Table in
§ 71.71(c)(1), which is most unfavorable
for the feature under consideration; and
(3) The initial internal pressure
within the containment system must be
the maximum normal operating
pressure, unless a lower internal
pressure, consistent with the ambient
temperature assumed to precede and
follow the tests, is more unfavorable.
*
*
*
*
*
§ 71.77
■
[Removed and Reserved]
15. Remove and reserve § 71.77.
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Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / Proposed Rules
§ 71.95
[Amended]
§ 71.106 Changes to quality assurance
program.
16. In § 71.95, remove paragraph
(a)(3).
■
§ 71.97
*
[Amended]
17. In § 71.97:
a. In the section heading, remove the
phrase ‘‘irradiated reactor fuel and’’;
■ b. In paragraph (b) introductory text,
remove the word ‘‘also’’;
■ c. In paragraph (d) introductory text
and paragraphs (d)(1) and (2), remove
the phrase ‘‘irradiated reactor fuel or’’;
and
■ d. In paragraph (f)(1), remove the
phrase ‘‘an irradiated reactor fuel or’’
and add in its place the word ‘‘a’’.
■
■
§ 71.100
[Amended]
18. In § 71.100(b), remove the
reference ‘‘71.77,’’.
■ 19. In § 71.106, revise the introductory
text of paragraph (b) to read as follows:
■
*
*
*
*
(b) Each quality assurance program
approval holder may change a
previously approved quality assurance
program without prior NRC approval, if
the change does not reduce the
commitments in the quality assurance
program previously approved by the
NRC. Changes to the quality assurance
program that do not reduce the
commitments shall be submitted to the
NRC every 24 months, in accordance
with § 71.1(a). If no changes were made
to the quality assurance program this
information shall also be submitted to
the NRC every 24 months, in accordance
with § 71.1(a). In addition to quality
assurance program changes involving
administrative improvements and
clarifications, spelling corrections, and
non-substantive changes to punctuation
or editorial items, the following changes
are not considered reductions in
commitment:
*
*
*
*
*
■ 20. In appendix A to part 71, in
paragraph V.b.:
■ a. In Table A–1, add the entries for Ba135m, Ge-69, Ir-193m, Ni-57, Sr-83, Tb149, and Tb-161 in alphanumeric order
and revise the entries for Ni-59, Rb(nat),
and Tb-157; and
■ b. In Table A–2, add the entries for Ba135m, Ge-69, Ir-193m, Ni-57, Sr-83, Tb149, and Tb-161 in alphanumeric order
and revise the entries for Ni-59, Tb-157,
Th(nat), and U(nat).
The additions and revisions read as
follows:
Appendix A to Part 71—Determination
of A1 and A2
*
*
*
*
*
V.b. * * *
TABLE A–1—A1 AND A2 VALUES FOR RADIONUCLIDES
Symbol of
radionuclide
Specific activity
Element and
atomic number
*
Ba-135m ..............
*
*
*
..............................
*
Ir-193m .................
*
..............................
*
Ni-57 ....................
Ni-59 ....................
*
Nickel (28) ...........
..............................
*
Rb(nat) .................
*
*
*
*
Terbium (65) ........
..............................
*
Tb-161 ..................
1.0 × 100 ..............
4.0 × 101 ..............
6.0 × 10¥1 ...........
Unlimited .............
*
..............................
*
*
*
*
*
5.4 × 102 ..............
2.7 × 101 ..............
1.1 × 103 ..............
1.6 × 101 ..............
Unlimited .............
*
Unlimited .............
..............................
Tb-149 ..................
Tb-157 ..................
2.0 × 101 ..............
*
..............................
Sr-83 ....................
A2 (Ci)b
A2 (TBq)
(TBq/g)
..............................
Ge-69 ...................
A1 (Ci)b
A1 (TBq)
1.0 × 100 ..............
8.0 × 10¥1 ...........
4.0 × 101 ..............
3.0 × 101 ..............
*
*
*
*
*
*
*
1.0 × 100 ..............
4.0 × 100 ..............
5.0 × 10¥1 ...........
Unlimited .............
*
Unlimited .............
*
6.0 × 10¥1 ...........
2.7 × 101 ..............
2.2 × 101 ..............
1.1 × 103 ..............
8.1 × 102 ..............
*
*
*
*
*
*
*
2.7 × 101 ..............
1.1 × 102 ..............
1.4 × 101 ..............
Unlimited .............
*
Unlimited .............
*
1.6 × 101 ..............
1.0 × 100 ..............
8.0 × 10¥1 ...........
4.0 × 101 ..............
7.0 × 10¥1 ...........
*
*
*
*
*
*
Unlimited .............
*
*
*
2.7 × 101 ..............
2.2 × 101 ..............
1.1 × 103 ..............
1.9 × 101 ..............
*
*
*
*
(Ci/g)
3.0 × 104 ..............
4.3 × 104 ..............
2.4 × 103 ..............
5.7 × 104 ..............
3.0 × 10¥3 ...........
6.7 × 10¥10 .........
4.3 × 104 ..............
1.9 × 105 ..............
5.6 × 10¥1 ...........
4.3 × 103 ..............
*
*
*
*
*
*
*
*
*
8.1 × 105
1.2 × 106
6.4 × 104
1.5 × 106
8.0 × 10¥2
1.8 × 10¥8
1.2 × 106
5.1 × 106
1.5 × 101
1.2 × 105
*
lotter on DSK11XQN23PROD with PROPOSALS1
TABLE A–2—EXEMPT MATERIAL ACTIVITY CONCENTRATIONS AND EXEMPT CONSIGNMENT ACTIVITY LIMITS FOR
RADIONUCLIDES
Activity
concentration for
exempt material
(Bq/g)
Activity
concentration for
exempt material
(Ci/g)
Activity limit
for exempt
consignment
(Bq)
Activity limit
for exempt
consignment
(Ci)
Symbol of
radionuclide
Element and
atomic number
*
Ba-135m ....................
*
...................................
*
*
*
1.0 × 102 ................... 2.7 × 10¥9 ................
*
1.0 × 106 ...................
2.7 × 10¥5
*
Ge-69 .........................
*
...................................
*
*
*
1.0 × 101 ................... 2.7 × 10¥10 ...............
*
1.0 × 106 ...................
2.7 × 10¥5
*
Ir-193m .......................
*
...................................
*
*
*
1.0 × 104 ................... 2.7 × 10¥7 ................
*
1.0 × 107 ...................
2.7 × 10¥4
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*
55734
Federal Register / Vol. 87, No. 175 / Monday, September 12, 2022 / Proposed Rules
TABLE A–2—EXEMPT MATERIAL ACTIVITY CONCENTRATIONS AND EXEMPT CONSIGNMENT ACTIVITY LIMITS FOR
RADIONUCLIDES—Continued
Activity
concentration for
exempt material
(Bq/g)
Activity
concentration for
exempt material
(Ci/g)
Activity limit
for exempt
consignment
(Bq)
Activity limit
for exempt
consignment
(Ci)
Symbol of
radionuclide
Element and
atomic number
*
Ni-57 ..........................
Ni-59 ..........................
*
Nickel (28) .................
...................................
*
*
*
1.0 × 101 ................... 2.7 × 10¥10 ...............
1.0 × 104 ................... 2.7 × 10¥7 ................
*
1.0 × 106 ...................
1.0 × 108 ...................
2.7 × 10¥5
2.7 × 10¥3
*
Sr-83 ..........................
*
...................................
*
*
*
1.0 × 101 ................... 2.7 × 10¥10 ...............
*
1.0 × 106 ...................
2.7 × 10¥5
*
Tb-149 ........................
Tb-157 ........................
*
Terbium (65) .............
...................................
*
*
*
1.0 × 101 ................... 2.7 × 10¥10 ...............
1.0 × 104 ................... 2.7 × 10¥7 ................
*
1.0 × 106 ...................
1.0 × 107 ...................
2.7 × 10¥5
2.7 × 10¥4
*
Tb-161 ........................
*
...................................
*
*
*
3.0 × 101 ................... 8.1 × 102 ...................
*
7.0 × 10¥1 ................
1.9 × 101
*
Th(nat) (b), (c) ...........
*
...................................
*
*
*
1.0 ............................. 2.7 × 10¥11 ...............
*
1.0 × 103 ...................
2.7 × 10¥8
*
U(nat) (b), (c) .............
*
...................................
*
*
*
1.0 ............................. 2.7 × 10¥11 ...............
*
1.0 × 103 ...................
2.7 × 10¥8
*
*
*
*
*
*
* * * * *
b Parent nuclides and their progeny included in secular equilibrium are listed as follows:
Sr-90 .......................
Zr-93 .......................
Zr-97 .......................
Ru-106 ....................
Ag-108m .................
Cs-137 ....................
Ce-144 ....................
Ba-140 ....................
Bi-212 ......................
Pb-210 ....................
Pb-212 ....................
Rn-222 ....................
Ra-223 ....................
Ra-224 ....................
Ra-226 ....................
Ra-228 ....................
Th-228 .....................
Th-229 .....................
Th-nat ......................
Th-234 .....................
U-230 ......................
U-232 ......................
U-235 ......................
U-238 ......................
U-nat .......................
Np-237 ....................
Am-242m ................
Am-243 ...................
Y–90
Nb-93m
Nb-97
Rh-106
Ag-108
Ba-137m
Pr-144
La-140
Tl-208 (0.36), Po-212 (0.64)
Bi-210, Po-210
Bi-212, Tl-208 (0.36), Po-212 (0.64)
Po-218, Pb-214, Bi-214, Po-214
Rn-219, Po-215, Pb-211, Bi-211, Tl-207
Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Rn-222, Po-218, Pb-214, Bi-214, Po-214, Pb-210, Bi-210, Po-210
Ac-228
Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212(0.64)
Ra-225, Ac-225, Fr-221, At-217, Bi-213, Po-213, Pb-209
Ra-228, Ac-228, Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Pa-234m
Th-226, Ra-222, Rn-218, Po-214
Th-228, Ra-224, Rn-220, Po-216, Pb-212, Bi-212, Tl-208 (0.36), Po-212 (0.64)
Th-231
Th-234, Pa-234m
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
Pa-233
Am-242
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.
* * * * *
lotter on DSK11XQN23PROD with PROPOSALS1
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
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]]>Agencies
[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
________________________________________________________________________
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.
========================================================================
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 https://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
