DRC-2023-072870 - 0901a068812831a0

DRC-2023-072870 195 North 1950 West • Salt Lake City, UT Mailing Address: P.O. Box 144880 • Salt Lake City, UT 84114-4880 Telephone [phone] • Fax [phone] • T.D.D. 711
www.deq.utah.gov Printed on 100% recycled paper
State of Utah
SPENCER J. COX Governor DEIDRE HENDERSON Lieutenant Governor
Department of Environmental Quality
Kimberly D. Shelley Executive Director DIVISION OF WASTE MANAGEMENT AND RADIATION CONTROL Douglas J. Hansen Director October 5, 2023
David Frydenlund, Executive Vice President, Chief Legal Officer Energy Fuels Resources (USA), Inc. 225 Union Boulevard, Suite 600
Lakewood, CO 80228 RE: Opportunity to Respond to SERP Process Comments Dear Mr. Frydenlund:
The Division of Waste Management and Radiation Control (Division) is in receipt of a letter from Uranium Watch, dated August 22, 2023 (DRC-2023-070710, copy attached). The Division would like to provide Energy Fuels Resources, Inc. (EFRI) with an opportunity to provide the Division with an appropriate
response to these comments. The Letter includes a number of comments concerning the White Mesa Mill License Condition 9.4 governing the Safety and Environmental Review Panel (SERP) process. The Division’s administrative record regarding this condition is somewhat limited. The Division understands that the Nuclear Regulatory Commission (NRC) added the SERP license condition (Amendment 3) before the State of Utah obtained Agreement State status for regulation of 11e.(2) byproduct material from the NRC.
The Uranium Watch comments are focused on legal issues. As a result, we have asked the Attorney General’s Office to assist in the Division’s review of these comments. The Division requests that EFRI provide comments within 45 days of the date of this letter.
If you have any questions, please contact Stevie Norcross by email at [e-mail] or by phone at 385-499-0511.
Sincerely,
Douglas J. Hansen, Director
Division of Waste Management and Radiation Control (Over)

DJH/SN/jk Enclosure: Uranium Watch, Information Comments, letter dated August 22, 2023 (DRC-2023-070710)
c: Mark Chalmers, President and CEO, Energy Fuels Resources (USA), Inc. (Email) David Frydenlund, Executive Vice President, Chief Legal Officer, Energy Fuels Resources (USA), Inc. (Email and Hard Copy) Michael Zody, Attorney at Law, Parsons Behle & Latimer (Email)
Kimberly D. Shelley, Executive Director, Utah Department of Environmental Quality (UDEQ) Ty L. Howard, Deputy Director, UDEQ Stevie Norcross, PhD, Asst. Director, Division of Waste Management and Radiation Control, UDEQ Phil Goble, Uranium Mills and Radioactive Materials Manager, Division of Waste Management and Radiation Control, UDEQ
Bret F. Randall, Assistant Attorney General, Utah Attorney General’s Office
Uranium Watch
P.O. Box 1112
Moab, Utah 84532
435-26O-8384
August 22, 2023
via electronic mail
Doug Hansen, Director
Division of Waste Management and Radiation Control
Utah Department of Environmental Quality
P.O. Box 144880
Salt Lake City, Utah 84114-4880
[e-mail]
Re: Advance Notice: Proposed Draft of New Version of R313-24 - Information
Comments
Dear Mr. Hansen:
Below are Uranium Watch’s Information Comments regarding the Division of Waste
Management and Radiation Control (DWMRC, or Division) Proposed Draft of New
Version of changes to Utah Rule R313-24. These Comments supplement Uranium
Watch’s April 25 and June 20, 2023, comments.
COMMENTS
1. Uranium Watch’s June 20, 2023, Comments discussed the use of the Safety and
Environmental Review Panel (SERP) process in place of a License Amendment Process:
The License Amendment Process includes license amendment applications,
Environmental Reports, Environmental Impact Analyses, public comments and hearings,
and amendments to existing licenses and license conditions. The June 20 Comments
stated: “The SERP process is conducted outside of the DWMRC statutory and regulatory
requirements.” The comments herein provide additional support for that assertion.
2. The DWMRC does not have any regulations related to the SERP process. The SERP
process is not based on any Nuclear Regulatory Commission (NRC) regulations
applicable to uranium mills and other uranium recovery operations under 10 C.F.R Part
40.
DRC-2023-070710
D. Hansen/DWMRC
August 22, 2023
2
3. A recent NRC document, “Safety Evaluation Report for Rare Element Resources Inc.
Proposed Demonstration Plant, Upton, Wyoming,” discusses the use of the SERP 1
process for source material licenses:
7.5 Safety and Environmental Review Panel (SERP) (page 24)
Regulatory Requirements
There are no regulatory requirements specifically related to source material
licensees using SERP to review and determine whether or not a license
amendment is required to change a process or procedure or conduct a test or
experiment in Title 10 of the Code of Federal Regulations. However, SERPs for
making these types of evaluations are based on the language in 10 CFR 50.59
(used by nuclear power reactor licensees) to perform evaluations for the conduct
of tests, changes to processes or procedures, or experiments to determine if those
activities can be conducted without the need for an NRC license amendment. As
part of the NRC’s initiative to transition to a performance based, risk-informed
inspection and licensing model detailed in SECY-98-144 (NRC,1998), the NRC
staff has adopted sections of 10 CFR 50.59 language for source and special
nuclear materials users and incorporated this flexibility to evaluate the need for
license amendments into source and special nuclear material licenses
using license condition language.
4. The requirements of 10 C.F.R. § 50.59 (Changes, Tests and Experiments) apply to
nuclear reactors and independent spent fuel storage facilities; they do not apply to NRC
10 C.F.R. Part 40 licensees, or Agreement State uranium mill licensees. The Rule only
references Parts 50 and 72. 2
When the NRC adopted 10 C.F.R. 50.59, there was no mention of possible application of
the rule to Part 40 licensees. The Final Rule states: “The Nuclear Regulatory
Commission (NRC) is amending its regulations concerning the authority for licensees of
production or utilization facilities, such as nuclear reactors, and independent spent fuel
storage facilities, and for certificate holders for spent fuel storage casks, to make changes
to the facility or procedures, or to conduct tests or experiments, without prior NRC
approval.” The Rule was meant to apply to types of nuclear fuel chain operations that are
very different from the processing of ore at licensed uranium mills and the disposal and
long-term care of the wastes from that processing.
Safety Evaluation Report for Rare Element Resources Inc. Proposed Demonstration Plant, 1
Upton, Wyoming, Materials License No. SUA-1603, Docket No. 040-38415, July 2023,
Accession No. ML23173A117.
<https://www.nrc.gov/docs/ML2317/ML23173A117.pdf>
64 Fed. Reg. 53582, 53582-53617; October 4, 1999. 10 CFR Parts 50 and 72, RIN 3150-AF94, 2
Changes, Tests, and Experiments, Nuclear Regulatory Commission; Final Rule. See Attached.
D. Hansen/DWMRC
August 22, 2023
3
The public was never given an opportunity to comment on the applicability of the
provisions of 10 C.F.R. § 50.59 to source material licensees, including uranium mill
licensees.
5. The Final Rule contains important definitions and applicability information, yet those
provisions only apply to Part 50 and Part 72 Licensees, not Part 40 licensees. The State of
Utah did not adopt 10 C.F.R. § 50.59 or changes to Part 72 into its regulations, because
the State does not, and cannot, have regulatory oversight of Part 50 or Part 72 facilities.
The SERP language used by the DWMRC and Utah Uranium Mill licensees does not
include some of the important definitions in the NRC Final Rule, thereby twisting the
purpose and intent of the NRC Rule.
6. The NRC developed specific Guidance to implement 10 C.F.R. § 50.59: “Guidance
for Implementation of 10 CFR 50.59, “Changes, Tests, and Experiments;” Regulatory
Guide 1.187, Rev. 3. This Reg. Guide applies to: 3
This RG applies to each holder of an operating license issued under 10
CFR Part 50, “Domestic Licensing of Production and Utilization Facilities” (Ref.
1), or a combined license issued under 10 CFR Part 52, “Licenses, Certifications,
and Approvals for Nuclear Power Plants” (Ref. 2), including the holder of a
license authorizing operation of a nuclear power reactor that has submitted the
certification of permanent cessation of operations required under 10 CFR 50.82(a)
(1) or 10 CFR 50.110 or a reactor licensee whose license has been amended to
allow possession of nuclear fuel but not operation of the facility.
The public did not have an opportunity to comment on the use of the Guidance for
facilities other than Part 50, 52, and 72 Production and Utilization Facilities. There is no
NRC Regulatory Guide regarding “Changes, Tests, and Experiments,” that applies to Part
40 uranium recovery or source material licensees.
7. In sum, there is no statutory or regulatory basis for the DWMRC’s use of the SERP
Process in place of a license amendment process for uranium mill and 11e.(2) byproduct
material licensees.
Thank you for providing the opportunity to comment.

GUIDANCE FOR IMPLEMENTATION OF 10 CFR 50.59, “CHANGES, TESTS, AND 3
EXPERIMENTS;” Reg. Guide 1.187, Revision 3, June 30, 2021. <https://www.nrc.gov/docs/
ML2110/ML21109A002.pdf>
D. Hansen/DWMRC
August 22, 2023
4

Sincerely,
/s/
Sarah Fields
Program Director
Attachments: As noted.
53582 Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
List of Subjects in 5 CFR Parts 831 and
842
Administrative practice and
procedure, Air traffic controllers,
Alimony, Claims, Disability benefits,
Firefighters, Government employees,
Income taxes, Intergovernmental
relations, Law enforcement officers,
Pensions, Reporting and recordkeeping
requirements, Retirement.
Office of Personnel Management.
Janice R. Lachance,
Director.
Accordingly, OPM is amending parts
831 and 842 of title 5, Code of Federal
Regulations, as follows:
PART 831—RETIREMENT
1. The authority citation for part 831
is revised to read as follows:
Authority: 5 U.S.C. 8347; §831.102 also
issued under 5 U.S.C. 8334; §831.106 also
issued under 5 U.S.C. 552a; §831.114 also
issued under 5 U.S.C. 8336(d)(2), Pub. L.
105±174, 112 Stat. 91; §831.201(b)(1) also
issued under 5 U.S.C. 8347(g); §831.201(b)(6)
also issued under 5 U.S.C. 7701(b)(2);
§831.201(g) also issued under sections
11202(f), 11232(e), and 11246(b) of Pub. L.
105±33, 111 Stat. 251; §831.204 also issued
under section 102(e) of Pub. L. 104±8, 109
Stat. 102, as amended by section 153 of Pub.
L. 104±134, 110 Stat. 1321; §831.303 also
issued under 5 U.S.C. 8334(d)(2); §831.502
also issued under 5 U.S.C. 8337; §831.502
also issued under section 1(3), E.O. 11228, 3
CFR 1964±1965 Comp.; §831.663 also issued
under 5 U.S.C. 8339(j) and (k)(2); §§831.663
and 831.664 also issued under section 11004
(c)(2) of Pub. L. 103±66, 107 Stat. 412;
§831.682 also issued under section 201(d) of
Pub. L. 99±251, 100 Stat. 23; subpart S also
issued under 5 U.S.C. 8345(k); subpart V also
issued under 5 U.S.C. 8343a and section 6001
of Pub. L. 100±203, 101 Stat. 1330±275;
§831.2203 also issued under section
7001(a)(4) of Pub. L. 101±508, 104 Stat.
1388±328.
Subpart A—Administration and
General Provisions
§831.108 [Removed]
2. Section 831.108 is removed.
3. In §831.114, paragraphs(b)(4) and
(c)(2)(iii) are revised to read as follows:
§831.114 Early retirement-major
reorganization, major reduction in force, or
major transfer of function
* * * * *
(b) * * *
(4) OPM may approve an agency's
request for voluntary early retirement
authority to cover the entire period of
the major reduction in force, major
reorganization, or major transfer of
function; or through the end of each
fiscal year, whichever is less.
(c) * * *
(2) * * *
(iii) The time period during which
voluntary early retirement will be
offered. At the agency's discretion, the
agency may request voluntary early
retirement authority to cover the entire
period of the major reduction in force,
major reorganization, or major transfer
of function; or through the end of the
fiscal year, whichever is less.
* * * * *
PART 842—FEDERAL EMPLOYEES
RETIREMENT SYSTEM—BASIC
ANNUITY
4. The authority citation for part 842
is revised to read as follows:
Authority: 5 U.S.C. 8461(g); §§842.104 and
842.106 also issued under 5 U.S.C. 8461(n);
§842.105 also issued under 5 U.S.C.
8402(c)(1) and 7701(b)(2); §842.106 also
issued under section 102(e) of Pub. L. 104±
8, 109 Stat. 102, as amended by section 153
of Pub. L. 104±134, 110 Stat. 1321; §842.107
also issued under sections 11202(f), 11232(e),
and 11246(b) of Pub. L. 105±33, 111 Stat.
251; §842.213 also issued under 5 U.S.C.
8414(b)(1)(B), Pub. L. 105±174, 112 Stat. 91;
§§842.604 and 842.611 also issued under 5
U.S.C. 8417; §842.607 also issued under 5
U.S.C. 8416 and 8417; §842.614 also issued
under 5 U.S.C. 8419; §842.615 also issued
under 5 U.S.C. 8418; §842.703 also issued
under section 7001(a)(4) of Pub. L. 101±508;
§842.707 also issued under section 6001 of
Pub. L. 100±203; §842.708 also issued under
section 4005 of Pub. L. 101±239 and section
7001 of Pub. L. 101±508; subpart H also
issued under 5 U.S.C. 1104.
Subpart B—Eligibility
§842.205 [Removed]
5. Section 842.205 is removed.
8. In §842.213,paragraphs (b)(4) and
(c)(2)(iii) are revised to read as follows:
§842.213 Early retirement-major
reorganization, major reduction in force, or
major transfer of function
* * * * *
(b) * * *
(4) OPM may approve an agency's
request for voluntary early retirement
authority to cover the entire period of
the major reduction in force, major
reorganization, or major transfer of
function; or through the end of each
fiscal year, whichever is less.
(c) * * *
(2) * * *
(iii) The time period during which
voluntary early retirement will be
offered. At the agency's discretion, the
agency may request voluntary early
retirement authority to cover the entire
period of the major reduction in force,
major reorganization, or major transfer
of function; or through the end of the
fiscal year, whichever is less.
* * * * *
[FR Doc. 99±25707 Filed 10±1±99; 8:45 am]
BILLING CODE 6325–01–U
NUCLEAR REGULATORY
COMMISSION
10 CFR Parts 50 and 72
RIN 3150–AF94
Changes, Tests, and Experiments
AGENCY: Nuclear Regulatory
Commission.
ACTION: Final rule.
SUMMARY: The Nuclear Regulatory
Commission (NRC) is amending its
regulations concerning the authority for
licensees of production or utilization
facilities, such as nuclear reactors, and
independent spent fuel storage facilities,
and for certificate holders for spent fuel
storage casks, to make changes to the
facility or procedures, or to conduct
tests or experiments, without prior NRC
approval. The final rule clarifies the
specific types of changes, tests, and
experiments conducted at a licensed
facility or by a certificate holder that
require evaluation, and revises the
criteria that licensees and certificate
holders must use to determine when
NRC approval is needed before such
changes, tests, or experiments can be
implemented. The final rule also adds
definitions for terms that have been
subject to differing interpretations, and
reorganizes the rule language for clarity.
Additionally, the final rule grants in
part and denies in part, a petition for
rulemaking (PRM±72±3) submitted by
Ms. Fawn Shillinglaw on December 9,
1995. This notice constitutes final NRC
action on this petition.
EFFECTIVE DATE: The amendments to
sections 72.3, 72.9, 72.24, 72.56, 72.70,
72.80, 72.86, 72.244, 72.246, 72.248 of
this rule are effective February 1, 2000.
Sections 50.59, 50.66, 50.71(e), and
50.90 become effective 90 days after
issuance of applicable regulatory
guidance. The NRC will publish a
document in the Federal Register that
announces the issuance of the
regulatory guidance and specifies that
the final rule becomes effective in 90
days. Section 72.212 and the
amendments to 72.48 are effective April
5, 2001.
FOR FURTHER INFORMATION CONTACT:
Eileen McKenna, Office of Nuclear
Reactor Regulation, U.S. Nuclear
Regulatory Commission, Washington,
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53583Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
DC 20555±0001, telephone (301) 415±
2189; e-mail: [e-mail].
SUPPLEMENTARY INFORMATION:
I. Background
II. Comments and resolution on proposed
rule topics
A. Organization of the rule requirements
B. Change to the facility as described in the
Safety Analysis Report
B.1 Definition of change
B.2 Definition of facility
C. Change to the procedures as described
in the safety analysis report
D. Tests and experiments not described in
the final safety analysis report
E. Safety analysis report
F. Minimal increase principle
G. Section 50.59(c)(2) criteria on increases
in probability or consequences
H. Possibility of an accident of a different
type from any previously evaluated in
the final safety analysis report (as
updated) is created
I. Possibility of a malfunction of a
structure, system, or component
important to safety with a different result
from any previously evaluated in the
final safety analysis report (as updated)
is created
J. Replacement criteria for ``margin of
safety as defined in the basis for any
technical specification is reduced''
K. Safety evaluation
L. Reporting and recordkeeping
requirements
M. No significant hazards consideration
determinations
N. Part 52 changes
O.1 Part 72 changes
O.2 Petition for Rulemaking (PRM±72±3)
O.3 Part 71 (Transportation) Comments
P. Other topics discussed in the notice and
comments not related to preceding topic
areas
Q Enforcement policy
R. Implementation
III. Section by section analysis
IV. Finding of no significant environmental
impact
V. Paperwork Reduction Act statement
VI. Regulatory analysis
VII. Regulatory Flexibility Certification
VIII. Backfit analysis
IX. Small Business Regulatory Enforcement
Fairness Act
X. National Technology Transfer and
Advancement Act
XI. Criminal penalties
XII. Compatibility of Agreement State
Regulations
List of Subjects
I. Background
The existing requirements governing
the authority of production and
utilization facility licensees to make
changes to their facilities and
procedures, or to conduct tests or
experiments, without prior NRC
approval are contained in 10 CFR 50.59.
Comparable provisions exist in §72.48
for licensees of facilities for the
independent storage of spent nuclear
fuel and high-level radioactive waste.
These regulations provide that licensees
may make changes to the facility or
procedures as described in the safety
analysis report (SAR), or conduct tests
or experiments not described in the
safety analysis report, without prior
Commission approval, unless the
proposed change, test, or experiment
involves a change to the Technical
Specifications (TS) incorporated in the
license or an unreviewed safety
question. Section 50.59(a)(2), as
codified, states the following:
A proposed change, test, or experiment
shall be deemed to involve an unreviewed
safety question (i) if the probability of
occurrence or the consequences of an
accident or malfunction of equipment
important to safety previously evaluated in
the safety analysis report may be increased;
or (ii) if a possibility for an accident or
malfunction of a different type than any
evaluated previously in the safety analysis
report may be created; or (iii) if the margin
of safety as defined in the basis for any
technical specification is reduced.
The rule also specifies recordkeeping
and reporting requirements associated
with such changes, tests, or
experiments.
Section 50.59 was promulgated in
1962 to allow licensees to make certain
changes that affect systems, structures,
components (SSC), or procedures
described in the SAR without prior
approval, provided certain conditions
were met. In 1968, the rule was revised
to modify some of the criteria for
determining whether prior NRC
approval was required. The intent of the
§50.59 process is to permit licensees to
make changes to the facility, provided
the changes maintain acceptable levels
of safety as documented in the SAR. The
process was thus structured around the
licensing approach of design basis
events (anticipated operational
occurrences and accidents), safety-
related mitigation systems, and
consequence calculations for the design
basis accidents.
On October 21, 1998 (63 FR 56098),
the NRC published a proposed rule to
revise §§50.59 and 72.48 to address a
number of issues concerning
implementation of the current rule, and
suitability of the criteria used to
determine when an unreviewed safety
question exists. Conforming changes
were proposed in other portions of the
regulations, including §§50.66, 50.71(e),
and 50.90 for production and utilization
facilities licensed under part 50.
Conforming changes were also proposed
in §72.212(b)(4).
The Commission proposed to make
similar changes to appendices A and B
of part 52, the standard design
certifications for the ABWR and CE
System 80+ designs respectively. These
regulations contain a change control
process similar to that in §50.59. As
noted in Section N, ``Part 52 changes''
below, the Commission has decided to
defer consideration of any changes to
part 52 until a later date.
In addition, the Commission proposed
to make parallel changes applicable to
independent spent fuel storage
installations (ISFSIs) licensed in
accordance with part 72. As part of the
proposed changes to part 72, the
Commission also proposed to extend the
change control authority granted to
ISFSI or monitored retrievable storage
(MRS) license holders (in §72.48) to
holders of NRC Certificates of
Compliance (CoC) for a spent fuel
storage cask design.
II. Comments and Resolution on
Proposed Rule Topics
The 60-day comment period for the
proposed rule closed on December 21,
1998. Comments were received from 60
organizations or individuals. Copies of
the comments are available for public
inspection and copying for a fee at the
Commission's Public Document Room,
located at 2120 L Street, NW.,
Washington DC. All comments were
considered in formulating the final rule.
The comments were submitted by 35
utilities with power reactor facilities; 2
representatives of nonpower reactor
licensees; 3 law firms representing
several utilities; 2 submittals from the
Nuclear Energy Institute (NEI); the U. S.
Enrichment Corporation; a nuclear
industry group; 6 nuclear utility
vendors, service companies or
consultants; 4 vendors or service
companies for spent fuel storage casks;
and 6 individuals. Forty commenters
endorsed (sometimes with further
comments) the NEI comments. NEI
stated in its comment letter that it
generally supports the Commission's
intent of the proposed rule but had a
number of comments or modifications
for certain specific provisions of the rule
that it wished the Commission to
consider in preparing the final rule. Of
those commenters who did not endorse
the NEI comments, most supported the
concept of the proposed rule, and made
recommendations to enhance or modify
certain elements of the rule. A few
commenters stated that the rule revision
was unnecessary and presented
supporting arguments. These
commenters felt that the Commission
should endorse NEI 96±07 ``Guidelines
for 10 CFR 50.59 Safety Evaluations,'' as
being sufficient to satisfy the existing
rule requirements. Many of the other
comments related to the content of
regulatory guidance, suggesting that
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53584 Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
examples be provided to amplify
particular points.
In the following sections, the NRC
presents a discussion and resolution of
the public comments, and the final
rulemaking language in a form that
parallels the order of discussion of
issues in the proposed rulemaking. The
organizational changes are discussed
first, followed by discussion of the
revised provisions in the rule. Although
the discussion of many of the topics
specifically focuses upon §50.59, these
matters are equally applicable to
§72.48, except as noted. Topics not
related to particular rule sections are at
the end of this discussion.
A. Organization of the Rule
Requirements
(1) Definitions
In the proposed rule, the Commission
added a new paragraph (a) to §50.59
that contains a number of definitions for
terms used in the rule. The Commission
sought comment on the need for
definitions as well as on the specific
definitions offered for the terminology.
Most commenters did not explicitly
address whether they thought
definitions were needed. One
commenter thought that adding
definitions only added confusion.
Another stated that although the terms
in the rule need to be defined, having
them in the rule means that any
subsequent changes in interpretation
would require rulemaking. The
Commission believes that having the
definitions in the rule adds clarity that
improves implementation of the rule,
and, in some cases, are necessary for
completeness of requirements.
Therefore the Commission has retained
several definitions in the final rule in
§§50.59(a) and 72.48(a). The specific
definitions are discussed in subsequent
sections.
(2) Applicability
The Commission proposed to place all
of the provisions concerning
applicability of the rule presently
contained in several subsections into
§50.59(b), which is clearly labeled
``Applicability.'' The rule applies to:
production and utilization facilities
(including power and non-power
reactors) that are authorized to operate,
and reactors (both power and non-
power) that have permanently ceased
operations. The few commenters who
addressed this topic were supportive of
this proposal. The final rule is
unchanged from the proposed rule in
this regard (except that §72.48 now
explicitly has a section with this
designation for consistency).
(3) Form of Prior Commission Approval
In the proposed rule, the Commission
combined §§50.59 (a) and (c) and
revised the regulation to state more
clearly that a licensee must apply for
and obtain a license amendment,
pursuant to §50.90, before
implementing changes, tests, or
experiments that involve either a change
to the TS or that satisfy any of the
criteria listed in new section 50.59(c)(2).
In addition, the Commission proposed
relocating an existing provision that
refers to changes to the TS not
associated with a change, test, or
experiment from §50.59 to §50.90.
Parallel changes to §72.48 and §72.56
were also proposed.
One aspect of the proposed rule that
drew comment concerned the
requirement to obtain a license
amendment before implementing a
change that involves a change to TS or
meets §50.59(c)(2) criteria. In particular,
for those instances in which a licensee
wishes to make a modification to the
facility, the use of which would require
a TS change (or meet one of the other
criteria), the commenters believe that it
is acceptable for a licensee to install and
test such a modification, as long as such
activities themselves do not place the
facility in a condition for which NRC
review is needed, and as long as the
modification is not actually used until
the amendment review has been
completed. These commenters believe
that waiting for NRC approval for use of
such modifications before beginning any
installation activity is unduly
restrictive. Typically this question arises
for plant modifications and installations
or complex engineering changes which
may take months or years to complete.
In the Commission's view, the
acceptability of such activities depends
upon the meaning of ``implementation''
and of which aspect of the change
requires NRC approval. If installing the
modification, or testing it after
installation would violate a TS, NRC
approval (of both the modification and
the revised TS) would be needed before
the change is implemented. In addition,
the licensee would need to determine
whether the test itself meets the criteria
in §50.59 so that prior NRC approval of
the test is not required. For changes that
are not inconsistent with existing TS,
but for which the licensee plans to
submit an amendment to later revise TS
to allow use of the modification (as for
instance a modification that may permit
less restrictive TS requirements),
proceeding with the installation, before
the approval is received, is at the
licensee's own risk with respect to
whether the Commission will approve
use of the modification. If the NRC finds
the proposed TS or the modification
unacceptable, the licensee would need
to appropriately revise the modification
or may be unable to reap the expected
benefits. If the licensee establishes that
installation and testing of a modification
do not require approval, but its use in
facility operations would, NRC approval
would be needed before the
modification could be put into effect.
With these clarifications, the
Commission accepts the comments on
this aspect. The final rule text is
unchanged from that offered in the
proposed rule.
(4) Criteria for Needing Commission
Approval of Changes, Tests, and
Experiments and Unreviewed Safety
Question (USQ) Designation
In the proposed rule, the Commission
proposed to remove the reference to the
term ``unreviewed safety question'' and
instead refer to the need to obtain a
license amendment. The Commission
concluded that this terminology has
sometimes led to confusion about the
purpose of the evaluation required by
§50.59. The purpose is to identify
possible changes that might affect the
basis for licensing the facility so that
any changes that might pose a safety
concern are reviewed by NRC to confirm
their safety before implementation. To
avoid confusion between a
determination of safety and a
determination of the need for NRC
approval, the Commission is removing
the term ``unreviewed safety question.''
In addition, the Commission proposed
to list the criteria (in the new
§50.59(c)(2)) that, if met, would require
prior Commission approval for a
proposed change, which would be in
the form of a license amendment. In the
proposed rule, the compound
statements contained within the
evaluation criteria of the current rule
were separated into several individual
criteria. The deletion of the term
``unreviewed safety question'' also
required a number of conforming
changes to other parts of the regulations.
Commenters generally supported
these proposed changes. A few
commenters stated that the
supplementary information should
explain that existing guidance referring
to ``USQ'' (such as Generic Letter 91±18,
Revision 1), is still applicable. Further,
commenters stated that a simple process
should be established by which licensee
technical specifications that use the
term ``USQ'' could be revised.
The Commission agrees that the term
USQ was used as a convenience to
describe those changes that met the rule
criteria for prior NRC review and
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1Under the NRC enforcement policy, §50.59 is
sometimes used to form the basis for a violation for
circumstances under which the as-built facility
differs from the FSAR, in that the existing condition
is a ``change'' from the ``as-described FSAR
condition'', and no evaluation was performed
supporting why the change could be made without
prior NRC approval. Such situations are referred to
as ``de facto'' changes.
approval, and that any guidance
referring to the same category of plant
changes is equally valid for describing
plant changes that would require prior
NRC review and approval under the
revised §50.59(c)(2).
The Commission considered the
merits of including specific language in
§50.59 that would address this point,
but ultimately did not include such
language for a number of reasons. First,
the NRC official record copy would not
be modified if licensees made changes
on their own (in accordance with the
rule language). Second, the intent of the
specific provision would be to permit
such changes; however, the fact that the
provision is contained in the rule may
make it a requirement to do so. This is
clearly an unintended consequence and
argues against including such language.
Finally, since there is no practical effect
of the wording as contained within the
TS, there is no compelling reason why
licensees would need to promptly
conform the wording of their TS. For
administrative convenience, the NRC
requests that upon such occasion as
those sections of the TS require NRC
approval for other reasons or a licensee
is requesting a license amendment in
some other area of the TS, the licensee
should include any necessary changes
to the existing TS language to bring the
plant-specific technical specifications
into conformance with the rule
language. Such changes could be made
at any time if a general formulation of
the requirement is used, as for example,
replacing ``USQ'' with ``requires NRC
approval pursuant to §50.59.'' Since
these are viewed as editorial changes
only, effectiveness of the existing TS is
not impacted. The implementation
period of the rule will give reasonable
opportunity to assure that the technical
specifications are appropriately
modified without the need to file a
separate amendment request.
(5) Changes in the Scope of the Rule
The Commission solicited public
comment on the need to revise the
scope of the rule in the notice for the
proposed rule. Specifically, the
Commission asked whether the scope of
the rule should be linked to the final
safety analysis report (FSAR), as
updated, or should the focus of the rule
be linked to another set of regulatory
requirements.
Only a few commenters indicated
interest in a redefinition of the scope of
the rule. These commenters suggested
that any attempt to redefine the scope of
the rule should be considered as part of
a longer term revision that might be part
of staff efforts to make the rule more risk
informed. Therefore, the NRC is not
revising the scope of the rule as part of
the final rule. The NRC will reconsider
the scope of the rule as part of its
ongoing initiatives to improve its
regulations to make them more risk
informed.
B. Change to the Facility as Described in
the Safety Analysis Report
In the proposed rule, the Commission
created a new §50.59(a) to contain
definitions for terms such as ``change''
and ``facility as described in the final
safety analysis report (as updated).'' The
definitions in §50.59 of ``change'' and of
``facility as described in the final safety
analysis report (as updated)'' were
written to more explicitly establish that
evaluation is required for changes to the
analyses and bases for the facility as
well as for physical or hardware
changes to the facility. The proposed
rule also explicitly stated that additions
were changes under the rule.
B.1 Definition of Change
In the proposed rule, the Commission
concluded that a ``change'' is a
modification of an existing provision
(e.g., structure, system, or component
design requirement, analysis method or
parameter), an addition or a removal
(physical removals or non-reliance on a
system to meet a requirement) to the
facility (or procedure) as described in
the FSAR.
Comment Summary: A number of
comments related to the definition of
change. The major topic areas of the
comments are summarized below. The
Commission's resolution of these
matters follows.
(a) Screening: Most of the commenters
were seeking revision of the definition
to allow screening of changes that
would not affect design functions. For
instance, some commenters, while
agreeing that additions should be
considered changes, also noted that
additions, if not limited by qualifiers
such as ``inconsistent with FSAR or
changing operation'', could mean that
even trivial additions to the facility or
to a procedure would require
evaluations. A few commenters thought
that additions should instead be treated
as ``tests or experiments,'' so that
evaluations would be needed only if the
additions were inconsistent with the
FSAR or outside the design basis.
(b) Replacement components or
maintenance: Other commenters sought
clarification as to whether particular
activities, such as the installation of
``equivalent'' components, or
maintenance activities are considered to
be changes requiring evaluation against
the criteria. For instance, replacement
equipment should only require review if
the replacement component has
characteristics that are different from
those described in the FSAR. For
maintenance, commenters stated that
taking SSC out of service for
maintenance is adequately covered by
maintenance rule requirements or TS,
and that a §50.59 evaluation should not
be required. Other commenters wanted
clarification that requirements for
environmental qualification of electrical
equipment were covered by §50.49,
such that equipment replacements that
are qualified per §50.49 are not
``reductions in margin of safety'' under
§50.59.
(c) Interdependent changes: A number
of comments concerned
``interdependent'' changes, that is,
under what circumstances can more
than one change be considered together
rather than individually. A few
commenters stated that the Commission
should adopt a position with respect to
interdependent changes that multiple
changes to the facility or its procedures
may be evaluated collectively if: (1)
They are interdependent as in the case
where a modification to a system or
component necessitates additional
changes to other systems or procedures
in order for the modified system to
perform its function or comply with its
design or licensing basis; (2) they are
performed collectively to address a
design or operational issue; or, (3) they
are otherwise planned as elements of a
single project undertaken to restore,
maintain or improve plant performance
or safety. Several commenters also
stated that examples would be helpful
to illustrate how closely related the
changes needed to be in order to be
viewed as interdependent.
(d) Removal: One commenter stated
that the term ``removal'' should be
clarified to include removal from
service, physical removal, retirement in
place, discontinued availability,
removal from the FSAR text or tables,
and removal from FSAR figures.
(e) De Facto Changes: One commenter
stated that the NRC should modify the
definition or other rule language to
explicitly state that the requirements
apply only to ``proposed'' changes and
not to so-called ``de facto'' changes.1
Another commenter thought the rule
language should explicitly codify the
resolution process under Generic Letter
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(GL) 91±18, by including language in
the rule such that the respective
requirements of Appendix B, criterion
16 and §50.59 do not interfere.
(f) Changes made in response to NRC
communications: Two commenters
asked if a proposed change that is the
direct result of a response to issues
raised in generic communications
requires evaluation under §50.59 to
determine the need for NRC approval, or
if it is already approved by the NRC.
The Commission notes that this subject
was also raised by NEI during a meeting
on guidance for minimal increases with
respect to changes being made to
conform with changes to regulations.
Resolution: The Commission has
modified the proposed rule language for
``change'' to be responsive to the issues
raised by these comments. In particular,
for comment (a), the Commission has
incorporated into the definition of
``change'' the phrase ``that affects design
function, method of performing or
controlling a function, or an evaluation
that demonstrates that intended
functions will be accomplished.'' The
Commission concluded that with this
revision, other comments about
``additions'' and ``removals'' have been
addressed (as for instance comment (d)).
The definition of change language will
allow licensees to eliminate the need to
further assess specific changes against
the criteria in the rule because the
nature of the change would never meet
the criteria of the rule and require prior
NRC review before implementation
(known in the industry as a screening
review). The capability to perform such
screening reviews for such minor
changes will reduce the burden of the
review process.
With respect to comment (b) about
whether specific types of activities are
``changes'', the Commission agrees that
clarification would be useful and will
work with affected stakeholders to
address the specific needs for regulatory
guidance to successfully implement the
final rule. In particular, the Commission
finds that guidance would be useful on
when ``replacement'' components must
be treated as a change, as for instance
because the replacement component has
characteristics different from those
described in the FSAR, compared to one
that is ``equivalent'' and thus not a
change. The Commission also agrees
that simply removing a component from
service for maintenance does not require
a §50.59 evaluation, but notes that
prolonged removal from service appears
indistinguishable in its effect from a
change that removes the component
from the facility. Further, there may be
circumstances under which
maintenance activities would place the
facility in a configuration not previously
considered, or require disabling of
barriers or movement of heavy loads to
accomplish. The Commission further
agrees that acceptability of
environmental qualification
requirements would be determined with
respect to §50.49. However, use of
different equipment would also require
a §50.59 review with respect to meeting
the evaluation criteria as now defined in
the rule (as discussed elsewhere, the
criterion on ``margin'' is being
removed). The Commission notes that
for certain changes, such as a change
that affects post-accident containment
conditions, although §50.49 may be the
applicable regulation for equipment
qualification, other aspects
(containment pressure) would need to
be evaluated under §50.59.
The Commission's previous
comments on interdependent changes
arises from concern that if multiple
changes were considered in a single
evaluation, certain aspects of the
``combined'' change could offset other
aspects and lead to a conclusion that the
set of changes did not require approval.
Certain of the other changes being made
to the final rule alleviate much of the
Commission's concern about this
practice. In particular, the Commission
has described in section J how changes
to methods, input parameters, and
facility changes should be evaluated in
determining whether the evaluation
criteria are met. Although the
Commission agrees with many of the
ideas offered by the commenters for
interdependent changes, the
Commission further believes that
providing further discussion and
examples in guidance on this point
would be useful.
The Commission did not modify the
rule language to specifically address
comment (e) on ``de facto'' changes or
GL 91±18 guidance, believing that
changes were not needed to allow the
process under GL 91±18 to be
implemented. The Commission did not
revise the rule language to specifically
state that ``changes'' resulting from
corrective actions under Appendix B do
not fall under the ``obtain amendment
prior to implementing'' requirement as
suggested by the commenter. The
Commission acknowledges that in those
instances of ``de facto'' changes, it is not
possible for the licensee to obtain NRC
approval prior to implementing a
change that has already occurred. In
these cases, the ``proposed change'' that
the licensee wishes to make is to its
FSAR such that it reflects the ``as-
found'' condition of the plant. The prior
approval specified in §50.59 is the
NRC's agreement with the resolution of
the nonconformance before the issue is
closed. For these instances, the
Commission views ``implementing the
change'' as meaning closeout of the
corrective action. Further, the
Commission does not plan to revise its
enforcement policy concerning de facto
changes (see also section Q below for
more discussion on enforcement for
§50.59).
With respect to item (f), the licensee
has an obligation to comply with the
regulations (including any changes), and
to respond appropriately to any generic
communication. The licensee must
examine the facility changes being made
to determine how the facility will
function with the change and identify
any potential impacts on safety. A rule
or generic communication may specify
a requirement to be satisfied, or the
nature of a change to meet a particular
intent, but rarely is the specific issue
presented at a level of detail necessary
for installation. For some facilities, or
some configurations, the ``generic''
solution intended by the rule or generic
communication may not achieve the
expected results, or there may be
alternative ways that would avoid other
problems. These issues can be pursued
in the licensee's response to the generic
communication or requirement.
The question about the need for NRC
approval for the specific means of
implementation of an action prompted
by NRC initiative (rule, order, or generic
communication) is less clear. As an
example, NRC has issued a rule
requiring the licensee to cope with a
station blackout. Suppose that the
means a licensee selects to meet the
requirement is to cross-connect a new
non-safety-related diesel to safety-
related buses. Before implementing this
modification, the licensee must evaluate
the change to determine whether the
particular method of satisfying the rule
has created other circumstances that
would warrant NRC review, such as if
the change would increase the
likelihood of malfunction of the buses.
Given these considerations, the NRC
concludes that changes made in
response to rules and generic
communications must be evaluated in
the same way as other changes a
licensee may wish to make, with the
conduct of §50.59 evaluations and
submittal of license amendment
requests as needed. Where there are
conflicts in requirements or schedules
resulting from these situations, the NRC
has an obligation to take timely and
appropriate action on the licensee's
submittals. To the extent that the
impacts of the generic communication
or rule are within the range of what the
NRC had considered in its deliberations
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on the rule or communication, the
approval of the licensee's submittal will
be straightforward.
In summary, the Commission has
included a definition of change as
meaning a modification or addition to,
or removal from the facility or
procedures that affects a design
function, method of performing or
controlling the function, or an
evaluation that demonstrates that
intended functions will be
accomplished. Other points raised by
the commenters, such as providing
examples, will be handled in the
regulatory guidance to be developed.
B.2 Definition of Facility
In the proposed rule, the Commission
concluded that changes to information
such as performance requirements,
methods of operation, the bases upon
which the requirements have been
established, and the evaluations should
be considered to constitute a change to
the ``facility as described in the FSAR
(as updated)''. The Commission
concludes that changes to methods and
other requirements in the FSAR, even if
not physical changes to the facility,
require evaluation under §50.59. If
changes to methods and performance
requirements were not so controlled, a
licensee might revise its analyses or
other information, update its FSAR, and
then subsequently conclude that a later
facility change does not require NRC
approval because the revised analysis or
acceptance requirement can still be
satisfied with the facility change (that
otherwise would have met the criteria as
requiring approval). Thus, the proposed
definition specifically itemized these
points.
Comment Summary: A few
commenters stated that it should be
clarified that changes, whether to
analysis methods or to the physical
facility, are only subject to §50.59
requirements if they are described in the
FSAR. Other commenters stated that if
the level of discussion within the FSAR
is unaffected by the change, there
should be no need for an evaluation.
NEI (as endorsed by other
commenters) stated that ``methods of
operation'' should be removed from the
definition of facility, as this was better
suited to the definition of ``procedures.''
Some commenters also were
concerned that the phrase ``required to
be included in the FSAR'' used in the
definition of facility was an attempt to
require licensees to look beyond the
FSAR, or to undertake actions to add
information to its FSAR. These
commenters thought such matters were
better handled as part of agency actions
concerning guidance for updating
FSARs (see for instance, Draft
Regulatory Guide DG±1083 and NEI 98±
03, ``Guidelines for Updating Final
Safety Analysis Reports'' ).
The Commission had included these
words in the rule as an attempt to limit
what part of the FSAR needed to be
considered for purposes of §50.59
evaluations. If information was not
required to be in the FSAR, then as
discussed under NEI 98±03, it could be
removed from the FSAR. On the other
hand, a licensee may wish to retain such
information in its FSAR for purposes of
completeness; then this part of the
definition would allow the licensee to
screen out changes to the information
that does not meet the definition of
facility as described. In view of the
confusion surrounding this phrase, and
in light of other proposed changes to
these definitions, the Commission has
deleted this phrase from the final rule.
A commenter stated that such
administrative changes as organizational
information, reporting relationships,
and job titles should be excluded from
the scope of §50.59.
Resolution: The Commission
considered these comments in selecting
the language that allows screening as to
whether a change to the facility affects
the content of the FSAR. As previously
noted in implementation guidance,
some SSC or subcomponents may not be
explicitly described in the FSAR, but
they have the potential to affect the
function of an SSC that is described.
The approach chosen by the
Commission for defining ``change'' as
relating to those additions,
modifications, and removals that affect
functions, methods of performing or
controlling functions and evaluation
methods also accomplishes an
important purpose for these issues.
Some changes a licensee may wish to
make to a component or procedure
could affect the functions or
performance requirements of other SSC.
Depending upon the level of detail
contained in the FSAR, the particular
component being changed may not be
explicitly described. If a modification to
that (non-described) component could
affect any SSC design function or
performance requirements that are
described, that modification affects the
design function, and thus is a change as
defined by §50.59(a) and thus requires
evaluation under §50.59. For example,
the bearings on a pump may not be
specifically mentioned or described in
the FSAR. However, the pump function
and performance requirement is
described. A change being made to the
bearings would need to be evaluated to
determine if it affects the function or
performance requirements of the pump,
and if so, whether the criteria in 50.59
(c) are met.
Changes to the definition of ``facility''
were made in response to the concerns
noted above from the commenters, such
as deletion of the phrases ``required to
be included ***,'' and ``methods of
operation.'' The Commission has
retained ``methods of evaluation'' as
being within the definition of ``facility,''
and as discussed under a later section,
added an evaluation criterion
specifically designed to provide a
standard for evaluation of such changes.
The Commission believes that the
definitions provided in the rule for
facility and procedures exclude the
indicated administrative type of changes
from §50.59, and further notes that
many of these details would be part of
a licensee's quality assurance plan that
is governed by the requirements of
§50.54(a), and therefore excluded from
the purview of §50.59 by virtue of
§50.59(c)(4).
The definition of facility includes
performance requirements and
evaluations included in the FSAR
which demonstrate that functions will
be accomplished. In part 54,
``Requirements for Renewal of Operating
Licenses for Nuclear Power Plants,''
§54.21(d) states that each renewal
application must contain an FSAR
supplement that contains a summary
description of the programs and
activities for managing the effects of
aging and the evaluation of time-limited
aging analyses for the period of
extended operation. As discussed in the
Statement of Considerations for the final
part 54, inclusion of the program
descriptions and analyses in the FSAR
provides the appropriate regulatory
oversight such that subsequent changes
are controlled by §50.59. The
Commission concludes that these
summary descriptions fall within the
definition of ``facility'' as demonstrating
that functions will be accomplished in
light of potential aging effects from the
period of extended operation. Therefore
changes that affect this information
require evaluation under §50.59. The
Commission further finds that
supplemental guidance or examples for
implementation specific to part 54
would be beneficial and NRC intends to
consider this as part of regulatory
guidance.
C. Change to the Procedures as
Described in the Safety Analysis Report
The Commission also proposed a
definition of ``procedures as described
in the safety analysis report'' in order to
have definitions in the rule for all the
major terms and criteria. This definition
includes the evaluations demonstrating
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that requirements are met, such as
assumed operator actions and response
times.
Commenters on the definition
primarily expressed concern with the
phrase ``conduct of operations'' because
licensees were concerned that this
language would inappropriately bring
administrative procedures within the
scope of the rule. Other commenters
suggested wording changes to clarify the
definition.
The Commission has decided to
remove the phrase ``conduct of
operations'' from the definition. The
Commission agrees that administrative
procedures are not intended to be
within the scope of the rule, and has
made other minor wording changes to
the final rule for clarity.
Changes Governed by Other Regulatory
Processes
In the proposed rule, the Commission
proposed to exclude from the scope of
§50.59 review, specific types of changes
to procedures where other requirements
and criteria have been established by
regulation for controlling these changes,
through a proposed provision in
§50.59(c)(1).
Commenters supported this proposal,
and suggested it be clarified to also refer
to plant changes in addition to
procedure changes. As an example,
emergency response facilities are
considered as part of the emergency
plans that are subject to §50.54(q). If
also described in the FSAR, there is a
potential for confusion as to whether
both a §50.54(q) and §50.59 evaluation
would be needed for a change to an
emergency response facility.
The Commission revised the rule
language to make the requested
clarification. Further, this section was
relocated to new §50.59(c)(4) in the
final rule. This language refers to
situations, such as §§50.54(a) and
50.54(q), where the regulations
explicitly define how changes are to be
reviewed, documented, and reported;
and thus, where a §50.59 evaluation
would be duplicative. Another example
would be §50.46, which establishes
criteria for reporting and for action for
changes involving methods for loss-of-
coolant analyses. A specific list of
regulations was not included in the rule
so that if other such rule sections
become available, §50.59 would not
need to be revised. The §50.59
obligation can only be replaced in
situations in which other rule
requirements specify the governing
change process, in order to prevent
duplication of reviews, not as a means
of avoiding change control
requirements.
A few commenters stated that
clarification should be included
concerning applicability of §50.59 for
certain documents controlled by a
variety of processes (e.g., Core Operating
Limit Reports contained in TS;
Technical Requirements Manual and
other matters (e.g., offsite dose
calculation manual (ODCM)) that have
been relocated from TS to other
controlled documents such as the FSAR;
and vendor topical reports, etc.).
The Commission notes that in NEI
98±03, which the NRC has proposed to
endorse through a regulatory guide,
there is discussion about incorporation
by reference of other documents (such
as ODCM, fire protection plan, etc) into
the FSAR. As discussed in Generic
Letter 86±10, ``Implementation of Fire
Protection Requirements,'' licensees
were encouraged to consolidate their
fire protection program documents and
incorporate them by reference into the
FSAR. Then, by the terms of a modified
license condition, licensees could make
changes to their fire protection program.
The vast majority of licensees have
made this change so that the program
description is incorporated into the
FSAR and program changes can be
made without NRC approval provided
the changes do not adversely affect the
ability to achieve and maintain safe
shutdown in the event of a fire (or
require an exemption). The Commission
sees no need to provide additional
clarification as the processes for control
of most of these documents are already
defined.
D. Tests and Experiments Not Described
in the Safety Analysis Report
The Commission proposed a
definition for ``tests and experiments
not described in the final safety analysis
report (as updated)'' to be included in
§50.59. The intent of the requirement is
that tests that put the facility in a
situation that has not previously been
evaluated or that could affect the
capability of SSC to perform their
intended functions should be evaluated
before they are conducted. Thus, the
definition focused upon the facility
being outside its design basis values or
inconsistent with the safety analyses in
the FSAR.
A few comments were made on this
topic, with some indicating that a
definition was not needed, and with
some noting that certain terms were
unclear or stating that the term
``activity'' should be used instead of
condition, to avoid confusion between
planned tests and identification of
degraded or nonconforming conditions.
(Note: because of administrative error,
the proposed rule text used the term
``condition,'' although in the proposed
rule supplementary information, the
term used was ``activity.'')
The Commission agrees with the
commenters and has used ``activity'' in
the final rule. Further, the Commission
believes that the phrase ``reactor, or any
of its structures, systems or
components'' is sufficiently clear to
reflect the intent that the determination
as to whether the activity is a test not
described in the FSAR, is not affected
by whether it is limited to only one
component, or involves a wider set, up
to and including the entire facility.
Therefore, the final rule has been
revised to contain a definition of ``test
or experiment not described in the final
safety analysis report (as updated)''
which has minor changes from the
definition offered in the proposed rule.
E. Safety Analysis Report
The Commission proposed to revise
the rule language to add a definition of
the ``final safety analysis report (as
updated)'' and to clarify in the
evaluation criteria that evaluations need
to account for changes made through
other processes that have not yet been
included in an update to the FSAR.
Thus, each of the evaluation criteria
contained a phrase referring to
evaluations and analyses performed
since the last FSAR update was
submitted. The rule referred to FSAR (as
updated), rather than to updated FSAR
to account for both non-power reactors
who are not required to submit updates
to their FSARs, and to any reactors
between the time of initial licensing and
the first required update. The definition
also refers to Final Hazards Summary
Report, because a few facilities were
licensed before the rules were revised to
require submittal of FSARs.
Commenters generally supported the
idea that the FSAR changes since the
last update submittal needed to be
considered in the §50.59 evaluations,
but sought clarification on a few details.
Further, commenters thought the rule
language could be simplified by
defining in one place that ``FSAR (as
updated)'' includes such information,
rather than including in each evaluation
criterion the phrase ``or in evaluations
performed pursuant to this section and
safety analyses performed pursuant to
§50.90 after the last final safety analysis
report was updated pursuant to §50.71
of this part.''
The Commission has modified the
rule text in response to these comments
by adding a new paragraph (c)(3) to
explicitly state that the ``FSAR (as
updated)'' for purposes of implementing
this paragraph, also includes the FSAR
update pages resulting from analyses
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and evaluations performed since the last
update was submitted. Accordingly, the
statements of the individual evaluation
criterion have been simplified.
Two commenters were concerned that
the requirement to consider other
evaluations since the last update
submittal would require a review of all
past evaluations to find the most
conservative result as the baseline for
these evaluations.
The Commission does not believe that
the rule requires such action. The
Commission's intent in stating that for
purposes of implementation of §50.59,
the FSAR (as updated) is considered to
include FSAR changes resulting from
evaluations of changes made since the
FSAR update is to ensure that decisions
about particular changes are made with
the most complete and accurate
information. If other changes did not
impact upon the accuracy of the FSAR,
they would not need to be examined. If
as a result of other changes, the licensee
will need to revise the FSAR at the next
update because the present information
is no longer accurate following that
change, that information may be
relevant to evaluation of a future change
that involves that part of the FSAR.
Indeed, for nonpower reactors, this
process has already been necessary
because these facilities are not required
to submit updates to their safety
analysis report. Nevertheless, they must
ensure that proposed changes are judged
with respect to the existing facility, not
the facility as originally described in the
FSAR at time of licensing. This
requirement does not make these
evaluations part of the updated FSAR
pursuant to §50.71(e); that rule requires
that the FSAR be updated to reflect the
effects of the changes and evaluations,
not that the evaluations themselves
become part of the updated FSAR.
Rather, the intent of the requirement is
that the changes that were the subject of
these evaluations be considered in the
process of determining what the
``facility as described'' now is such that
the reference for subsequent evaluations
is complete and accurate.
One commenter stated that it should
be made clear that the FSAR (as
updated) includes the TS and bases
because these documents sometimes
contain information, such as applicable
operating modes, not in the FSAR that
is relevant to the evaluation process. A
few other commenters thought the
definition for ``FSAR'' should include
other documents such as staff safety
evaluations, selected commitments and
other licensing documents.
The Commission does not agree that
these documents fall within the
required scope of the rule, or that they
are part of the FSAR. However, as noted
in existing guidance, licensees are free
to refer to other documents to assist in
understanding the implications of the
change, but the rule language does not
require such reviews.
F. Minimal Increase Principle
Strict interpretation of the existing
rule language related to the probability
of an accident or a malfunction has lead
to significant burden to the industry
with no clear safety benefits. Therefore,
in the proposed rule, the Commission
relaxed the standard for which prior
NRC review would be required by
revising existing paragraph
§50.59(a)(2)(i) of the rule. The specific
proposal was to replace the phrase ``may
be increased'' with ``would result in
more than a minimal increase.'' As
previously discussed, the present
§50.59(a)(2)(i) is being expanded into
four separate criteria, two for occurrence
of accidents and malfunctions and two
for consequences.
The information that can be revised
under §50.59 is limited to that which
does not require review under any other
sections of the regulations; thus, it is
information is of less direct importance
to public health and safety. In
consideration of the conservatisms in
NRC design and analysis requirements
and acceptance criteria, ``minimal''
variations in probability of occurrence
or consequences of accidents and
malfunctions should not affect the basis
for the previous licensing decision.
During the plant licensing process,
accident probabilities were assessed in
relative frequencies (such as likely to
occur more than once, likely to occur
once during the life of the plant, or
limiting fault that is not likely to occur
during the life of the plant). System
train and equipment failures were
generally postulated to gauge the
robustness of the design, without
estimating their likelihood of
occurrence. In this light, minimal
increases in probability would not
significantly change the licensing basis
of the facility and could not impact the
conclusions reached about acceptability
of the facility design.
Further, the limits for radiological
consequences established in the
regulations and in the Standard Review
Plan are conservatively chosen, so that
minimal increases also would not
impact the safety determination if
demonstrated by a suitably conservative
analysis. The Commission therefore
concluded that the proposed criteria
would provide reasonable assurance
that those changes that would affect the
NRC's basis for licensing would be
identified as requiring NRC approval
before implementation. The proposed
revisions to the §50.59 criteria would
provide some degree of flexibility for
licensees to make changes with smaller
impacts without the need to obtain a
license amendment.
On the other hand, the Commission
intends to limit the amount of increase
in probability or consequences of
accidents such that it remains
substantially less than a ``significant
increase'' as referred to in §50.92. In
accordance with §50.92, a license
amendment involving a significant
increase in the probability or
consequences of an accident previously
evaluated would be categorized as a
``significant hazards considerations''
and any hearing must be completed
prior to issuance of the amendment.
Although the final rule allows
minimal increases, licensees still must
meet applicable regulatory limits and
other acceptance criteria to which they
are committed (such as are contained in
Regulatory Guides and nationally
recognized industry consensus
standards, e.g., the ASME B&PV Code
and IEEE Standards). Further,
departures from the design, fabrication,
construction, testing, and performance
requirements as outlined in the General
Design Criteria (appendix A to part 50)
are not compatible with a ``no more than
minimal increase'' standard. Because
the ``no more than minimal'' standard
allows for there to be some increase
compared to the current requirement,
which would have required any
increase to be submitted for prior staff
review, NRC needs to establish a point
beyond which one would conclude that
the increase is not minimal. Application
of the ``minimal increase'' concept to
the specific criteria in the revised final
rule is discussed in the next sections.
G. Section 50.59 (c)(2)Criteria on
Increases in Probability or
Consequences
For each of the four evaluation criteria
replacing existing §50.59(a)(i), the
Commission presented language in the
proposed rule reflecting the ``minimal
increase'' principle. Resolution of each
of these criteria is discussed below,
including consideration of the public
comments.
For each criterion proposed, the
Commission had presented guidance on
how the rule could be met, including
values as to when the Commission
would conclude that each revised
criterion is not met. Comments received
on this guidance are discussed below.
The Commission also notes that
regulatory guidance will be provided
that is derived from this discussion.
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As the rule provides a qualitative
standard of ``no more than minimal,''
quantitative calculations are not
required except for those instances in
which a licensee decides to offer
quantitative arguments as part of its
evaluation. This is expected to occur for
some instances involving increases in
consequences, where licensees may
perform calculations of the predicted
dose from postulated accidents.
(i) More Than a Minimal Increase in the
Frequency of Occurrence of an Accident
Previously Evaluated
For criterion (i), the final rule requires
prior NRC approval if the change results
in more than a minimal increase in the
frequency of occurrence of an accident
previously evaluated in the FSAR (as
updated). Several commenters agreed
with the premise that ``minimal''
increases in probability of accidents
should not require prior NRC approval.
No specific comments were received on
the rule language itself. Issues about
guidance are discussed below.
The only change made by the
Commission in the final rule language
from the proposed rule is the
substitution of ``frequency'' for
``probability.'' This was done to provide
a better representation of the attribute of
concern, that is, occurrence over some
period of time, and to emphasize that
what is of interest is whether the
proposed change has the effect of
making the accident occur more often.
Guidance for Frequency of Accidents
In the proposed rule, the Commission
offered guidance concerning ``minimal''
with respect to increases in probability
(now frequency). Several comments
were received on certain of these
statements, as noted below.
First, the Commission had noted that
the current guidance in NEI 96±07
stating: ``Where a change in probability
is so small or the uncertainties in
determining whether a change in
probability has occurred are such that it
cannot be reasonably concluded that the
probability has actually changed (i.e.
there is no clear trend towards
increasing the probability), the change
need not be considered an increase in
probability'' satisfies the proposed NRC
standard for increases in frequency of an
accident. Commenters agreed with the
characterization that this guidance
would satisfy the rule, but also noted
that the rule language provides more
flexibility than is presently afforded by
the NEI guidance.
Second, the Commission had stated
that in order to be considered as a
minimal increase, the resulting
frequency of occurrence (considering
the change, test, or experiment) must
still satisfy the event frequency
classification provided in the licensee's
FSAR (as updated). Typically, these
would be anticipated operational
occurrence (expected once a year) or
design basis accidents (not expected
during life of plant, but sufficiently
credible to require mitigation). The use
of frequency classifications will not
apply for all facilities subject to §§50.59
or 72.48, but is included here because
it was a consideration in the licensing
of most operating power plants. Some
commenters sought clarification as to
whether increases that remain within
the frequency classification would
satisfy the ``no more than minimal
increase'' criterion. Changes that result
in a change in classification do not meet
the standard; however, remaining
within the classification is not sufficient
to conclude that no more than a
minimal increase has occurred because
qualitative judgments are not as rigorous
as quantitative assessments and the
accident categories and their
uncertainties may be large. The
Commission agrees that the effect of the
change on the frequency of the accident
must be discernible and attributable to
the change in order to exceed the ``more
than minimal'' increase standard, as
compared to uncertainty about the
existing frequency value and how it
might be quantified.
Some commenters stated that the
``minimal increase in probability''
standard was too vague and sought more
explicit criteria. Others requested
quantitative standards for determining
minimal increases in probability, and in
particular, guidance for using risk
insights or probabilistic risk analysis to
determine when a more than minimal
increase in probability has occurred. For
instance, commenters thought that the
values for changes in core damage
frequency or large early release
frequency in Regulatory Guide (RG)
1.174, ``An Approach for Using
Probabilistic Risk Assessment in Risk-
Informed Decisions on Plant-Specific
Changes to the Licensing Basis,'' might
be used. However, this RG was
developed for the purpose of guiding
changes to the licensing basis where the
staff was reviewing and approving the
change, not for changes made under
§50.59. The Commission concludes that
if use is to be made of PRA in §50.59,
more fundamental changes to the rule
would be necessary to provide a
coherent set of requirements, in that
§50.59 deals with design basis events,
and RG 1.174 deals with risk including
that from severe accidents beyond the
design basis. In addition, RG 1.174 is
specifically dealing with operating
power reactors. Applicability to other
facilities would need to be examined.
The Commission acknowledges that it
may be possible to develop more
guidance that could be used in a
quantitative sense to judge minimal
increases. As part of development of the
guidance, the NRC will consider using
the values developed as part of the
revised oversight process (SECY±99±
07), so that if the resultant likelihood of
occurrence remains well within the
acceptable ranges given for initiating
events, that the increase is ``minimal.''
(ii) Minimal Increase in Likelihood of
Malfunction of Structures, Systems or
Components
In the proposed rule, §50.59(c)(2)(ii)
would require NRC approval for a
change that would result in ``more than
a minimal increase in the probability of
malfunction of equipment important to
safety previously evaluated in the FSAR
(as updated).'' Similar changes were
proposed in §72.48(c)(2)(ii), except for
use of the term ``structures, systems, and
components'' (SSCs) rather than
equipment. These differences in
wording reflected differences between
existing language in §§50.59 and 72.48.
Commenters supported the idea that
``minimal'' increases should not require
approval. Commenters also suggested
that the terminology in §§50.59 and
72.48 should be made more consistent
between the two sections.
In the final rule, the Commission has
revised the criterion in §50.59 by
referring to SSC rather than to
equipment. The Commission concludes
that the term ``SSC'' is commonly used
in both parts 50 and 72 and is well
understood, and that ``equipment'' was
an older term that does not have a
unique meaning requiring its use. For
the final rule, the Commission has also
substituted the term ``likelihood'' for
``probability.'' This change was made to
acknowledge that while the criterion
refers to ``minimal'' increases, the
Commission is not implying that
quantitative assessments are expected.
The Commission concludes that the
word ``likelihood'' is more generally
understood to represent qualitative
judgments.
Guidance for Likelihood of Occurrence
of Malfunction
In the proposed rule, the Commission
discussed the following positions as
guidance for implementing the criterion
of a ``more than minimal'' increase in
probability (now likelihood) of a
malfunction of equipment (now SSC).
First, the Commission noted that the
existing guidance in NEI 96±07 states:
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``Where a change in probability is so
small or the uncertainties in
determining whether a change in
probability has occurred are such that it
cannot be reasonably concluded that the
probability has actually changed (i.e.
there is no clear trend towards
increasing the probability), the change
need not be considered an increase in
probability.'' Continued use of this
guidance for a determination of whether
criterion (i) has been met is satisfactory.
Commenters agreed with this guidance,
but also believe that this does not
represent the outer bound of what
would be acceptable to meet the rule.
The Commission agrees with this
comment.
Second, the Commission concluded
that the likelihood of malfunction of
SSC important to safety previously
evaluated in the FSAR (as updated)
would not be more than minimally
increased if ``design bases'' assumptions
and requirements are still satisfied (i.e.,
the seismic or wind loadings,
qualification specifications, etc). Thus,
for instance, a change that would cause
piping stresses to exceed their code
allowable values would be more than a
minimal increase in likelihood of
malfunction. Commenters stated that if
design basis requirements are met, there
is no increase in probability. The
Commission agrees with the essence of
this comment, but was attempting to
help licensees comply with the rule
language by offering ways of
demonstrating that the criterion is
satisfied. Changes that would invalidate
specific commitments made for
redundancy, diversity, separation, and
other such design characteristics, would
be considered as ``more than a minimal
increase in likelihood of malfunction,''
and thus would require prior NRC
approval.
In the proposed rule, the Commission
stated that for purposes of determining
whether this criterion has been satisfied,
the probability of malfunction would be
no more than minimally increased if a
new failure mode as likely as existing
modes is introduced. Some commenters
indicated that the presence of new
failure modes should not be a
determinant as to whether probability of
malfunction has increased; rather, it is
whether the effects of the failure modes
have previously been considered that
would determine the need for NRC
review consistent with §50.59(c)(2)(vi).
The Commission finds that the question
of likelihood is not addressed if new
failure modes are only examined with
respect to criterion (vi), since that
criterion looks only at whether the
effects of the failure are bounded, not
how likely it is to occur. However, since
likelihood can be increased regardless of
whether new failure modes are
involved, the Commission has deleted
this statement as proposed guidance for
assessing increases in likelihood.
Additions of components to a system
(cabling, manual valves, protective
features) would not generally be viewed
as more than a minimal increase in
likelihood of malfunction, provided that
applicable design and quality standards
are followed. For example, adding
protective devices to breakers, or
installing an additional drain line (with
appropriate isolation capability) would
not be increases in likelihood of
malfunction. However, there could be
situations where such additions would
impact upon how a system performs its
functions that might not satisfy the
§50.59 criteria (for example, a cross-
connect between trains that is not
suitably isolated).
Substitution of one type of component
for another (as for instance, an air-
operated valve for a motor-operated
valve), would also be viewed as no more
than a minimal increase in likelihood of
malfunction, provided requirements for
redundant motive force, quality, and
other requirements are met (and of
course that any new failure modes are
already bounded by the analysis).
(iii) and (iv) Minimal Increases in
Consequences of Accident or
Malfunction
In the proposed rule, the Commission
revised the existing criterion concerning
increases in consequences from a
standard of ``may be increased'' to
``more than minimally increased,'' and
separated the two statements on
consequences within §50.59(a)(2)(i) into
separate criteria. Only a few comments
were received concerning the rule
language itself. One commenter stated
that the two criteria on consequences
should not be separate, since
consequences would only result from
accidents, and having another criterion
might force evaluators either to
duplicate their documentation, or
struggle to explain why consequences
were not increased for malfunctions.
The Commission concludes that having
separate criteria provides greater clarity
and is consistent with common practice.
Further, the criteria cover different
types of changes, that is, some that arise
from malfunctions (such as failure of a
waste tank or filter systems), and others
that might arise from changes in source
term or timing of mitigation systems,
that are more pertinent to ``accidents.''
Licensees may combine their responses
to questions and reference other
sections when preparing evaluations.
Commenters requested two areas of
clarification. First, they asked if
consequences refers only to radiological
consequences (dose), and second
whether consequences refers only to
those associated with accidents and not
from normal operations or anticipated
operational occurrences. The rule
reference to consequences is intended to
relate directly to radiological
consequences, and not to other
outcomes that are covered by the
remaining criteria. Secondly, the
Commission notes that 10 CFR part 20
establishes requirements for protection
against radiation during normal
operations. For anticipated occupational
occurrences, NRC requirements are such
that there should not be any radiological
consequences. However, the
Commission also wishes to clarify that
``consequences of accidents'' includes
not only offsite exposure, but also dose
to operators in the control room (in
accordance with General Design
Criterion 19 of appendix A to 10 CFR
part 50) or other onsite personnel,
resulting from accidents and
malfunctions previously evaluated in
the FSAR.
The language in the rule for criterion
(iii) was unchanged from the proposed
rule; for criterion (iv), the term
``systems, structures, or components''
was substituted for ``equipment'' as it
was for criterion (ii), for the reasons
already discussed.
Guidance for Minimal Increase in
Consequences
In the proposed rule, the Commission
had discussed several positions that
might be helpful in developing guidance
that would successfully implement the
revised rule. First, the Commission
agreed with the guidance in NEI 96±07
which states: ``Where a change in
consequences is so small or the
uncertainties in determining whether a
change in consequences has occurred
are such that it cannot be reasonably
concluded that the consequences have
actually changed (i.e., there is no clear
trend towards increasing the
consequences), the change need not be
considered an increase in
consequences.'' No specific comments
were received on this point.
Second, if a licensee has performed an
analysis with certain bounding
assumptions, and the change would
increase a specific parameter from its
present value to a different value that is
still bounded by the value assumed in
the analysis, the NRC concludes that
such a change satisfies the criterion of
``no more than a minimal increase in
consequences.'' In fact, as noted by
some of the comments, this is no
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2In the Standard Review Plan, NUREG±0800, the
NRC established acceptance criteria for certain
events that are considered of greater likelihood than
the limiting accidents as a small fraction of the part
100 guidelines. Thus, for instance, for a steam
generator tube rupture, the SRP guideline is that the
dose be 10 percent of the part 100 value. For the
postulated accident with an assumed preaccident
iodine spike in the reactor coolant at the time the
tube rupture occurs, the full part 100 value is the
acceptance criterion.
3GDC 19 requires adequate radiation protection
to permit access and occupancy of the control room
under accident conditions without personnel
receiving radiation exposure in excess of 5 rem
whole body or its equivalent to any part of the
body, for the duration of the accident. Part 100
establishes requirements for exclusion area and low
population zones around the reactor so that an
individual located at any point on its boundary
immediately following onset of the postulated
fission product release would not receive a total
radiation dose to the whole body in excess of 25
rem or a total radiation dose of 300 rem to the
thyroid for iodine exposure. For future applications,
as noted in subpart B to 10 CFR part 100, the
radiological consequences are to meet the criteria
stated in §50.34(a)(1), which sets a dose of 25 rem
total effective dose equivalent (TEDE).
increase in consequences, because the
bounding analysis is what determines
the value from which a change is being
judged.
Third, if a licensee would need to
change its design basis assumptions or
analytical methods, or both, to
demonstrate that the change in
consequences satisfies this guidance,
then the NRC does not view the change
as minimal and would expect the
licensee to submit a license amendment
for such a change. This position is
consistent with the logic presented as
the basis for implementing new
criterion §50.59(c)(2)(viii), which will
be discussed in greater detail below.
Some commenters thought that adopting
methodologies that have been approved
by NRC in certain contexts (such as use
of International Conference on
Radiation Protection (ICRP) dose
conversion factors, or credit for
suppression pool scrubbing) should be
allowable under §50.59. New criterion
(viii), discussed in section J below,
specifies under what conditions changes
to evaluation methods can be changed
without prior NRC approval.
In the proposed rule, the Commission
proposed a graduated approach,
consistent with the concept of
``minimal'' being small enough so as not
to impact the basis for the acceptability
of the previous licensing decision. The
Commission proposed that when the
facility is far from the limit, a larger
increase could be accommodated
without concern about impact on the
basis for acceptability. The Commission
did not believe that allowing increases
up to the regulatory values without
approval was consistent with a
``minimal'' increase standard, and was
not consistent with the purpose of the
rule, that is, to allow the NRC the
opportunity to confirm the adequacy of
the licensee's review of the change
before it is implemented.
The proposed rule offered three
different ways to define what would
constitute a minimal increase in
consequences. Most commenters
favored the third method (10% of the
difference between the calculated value
and the regulatory guidelines) over the
other two. Other commenters thought
the limits themselves should be the
point at which NRC review would be
needed, or offered other suggestions,
such as allowing 20 percent of the
difference. Comments were also
received about the use of Standard
Review Plan guideline values 2 as they
are not in the regulations and that for
some plants, the existing analysis may
exceed the guideline such that no
changes would be allowed. Some
commenters also expressed concern
about the criterion for those situations
where a previous change may have
resulted in a decrease in consequences,
and a subsequent change that increased
consequences would exceed the 10
percent difference, but would not have
done so if the first change had not
occurred.
During the comment period, some
commenters were concerned that as the
rule is currently planned to be
implemented, they would have no
flexibility under the rule if their
calculated consequence values were
already in excess of the current SRP
guidelines. In general, the Commission
agrees that for cases where a licensee is
licensed with calculated consequences
in excess of the established SRP
guidelines, only limited flexibility
under this provision of the revised rule
would exist for changes that increased
the calculated radiological
consequences of accidents. In this
regard, the Commission does view
differences of about 0.1 rem as being
within the error or uncertainty of design
basis-type radiological consequences
analysis such that NRC review of such
changes is not needed.
The Commission has taken these
comments into account in revising the
``minimal'' increases in consequences
aspects of the final rule. The
Commission will conclude that the
requirements of the rule are met if the
calculated doses from a change at a
facility would be less than 10 percent of
the remaining margin between current
calculated dose values and acceptance
values in the regulations3 (e.g., GDC 19
or part 100) for the particular accident.
Under this approach, the threshold for
what constitutes a minimal change
varies as a licensee approaches the
regulatory limit. The amount of change
allowed would decrease as the limit is
approached, and the limit could not be
exceeded without prior NRC review.
Specifically, it is no more than a
minimal increase in consequences if the
increase is less than or equal to the more
limiting of either 10 percent of the
difference between the existing
calculated value and the regulatory
guideline value (10 CFR part 100 or
GDC 19 as applicable), or has reached
the SRP guideline value for the
particular design basis event.
Examples
The Commission has selected several
examples to illustrate the
implementation of this criterion. In each
example, the Commission assumes that
the calculated consequences do not
include changes in methodology. As
discussed later, changes in methodology
used to calculate radiological
consequences would fail new criterion
(viii) of the revised rule and require
prior NRC review regardless of how
small the increase would be in the
calculated radiological consequences.
Example 1 involves a case in which
a licensee has a calculated fuel handling
accident (FHA) dose of 50 rem to the
thyroid at the exclusion area boundary.
Because of some change in the facility,
the calculated FHA dose increases to 70
rem. Under the revised final rule, ten
percent of the difference between the
calculated value and the regulatory
limits is 25 rem (10% of 250). The SRP
acceptance guideline is 75 rem. Since
the calculated increase is less than 25
rem and the total is less than the SRP
acceptance guidelines, then the revised
§50.59 consequence criterion would not
trigger the need for a prior NRC review
and a licensee may make the change to
the facility.
Example 2 involves a case in which
the calculated consequences for a steam
generator tube rupture accident are 25
rem at the exclusion area boundary.
Because of a change in the plant, the
calculated consequences increase to 29
rem. The implementation of the revised
rule language would permit these
changes to occur because the new
calculated doses do not exceed the
established SRP acceptance criteria nor
does the incremental change in
consequences (4 rem) exceed 10 percent
of the difference between the previous
calculated value and the regulatory limit
of 300 rem. Ten percent of the
difference between the acceptance
criteria (300 rem) and the calculated
value (25) is 27.5 (10% of 275) rem;
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since 4 is less than 27.5, this change
satisfies the criterion.
Example 3 involves a case in which
the calculated consequences of a fuel
handling accident are 25 rem to the
thyroid at the exclusion area boundary.
Because of a proposed change in the
facility, the calculated consequences
increase to 65 rem. For this case, the
revised calculated consequences are still
less than the SRP acceptance guidelines
of 75 rem; however, the incremental
increase in consequences (40 rem)
exceeds the 10 percent of the difference
to the regulatory limit of 300 rem
(which would be 27.5 rem). For this
example, the change results in more
than a minimal increase in
consequences and thus requires NRC
approval pursuant to §50.59(c)(2)(iii).
If Example 3 had been an event for
which no SRP value was specifically
established, so that the part 100
guideline was the only applicable
standard, the rationale would be that an
increase up to 52.5 (25+27.5) rem would
meet the ``minimal increase'' criterion.
Example 4 involves a case where the
calculated dose to the control room
operators following a loss of coolant
accident is 4 rem whole body. A change
is made to the control room ventilation
system such that the calculated dose
increases to 4.5 rem. The regulations
dictate that the control room doses are
to be controlled to less than 5 rem by
General Design Criterion 19. Although
the new calculated doses are less than
the regulatory limits for the operators,
the incremental increase in dose (0.5
rem) exceeds the value of 10 percent of
the difference between the previously
calculated value and the regulatory
value (10% of 1 rem = 0.1 rem). This
change would require prior NRC review
before the licensee could implement the
change.
As an example of the ``calculational
error'' concept, suppose the existing
approved analysis for a fuel handling
accident at a plant predicts an offsite
dose to the thyroid of 77 rem. The SRP
acceptance guideline for this event is 75
rem. The change that a licensee wishes
to make would predict an increase in
the calculated dose from 77 to 77.1 rem.
In this case, the proposed change could
be made under §50.59 because the
calculated value, even though greater
than the SRP value, is satisfied within
the level of uncertainty specified above.
However, for this example, the
Commission notes that increases in
consequences that would increase the
calculated consequences to 77.2 rem
would require prior NRC review before
the specific change could be
implemented.
H. Possibility of an Accident of a
Different Type From Any Previously
Evaluated in the Final Safety Analysis
Report (as Updated) Is Created
The Commission had proposed that
the language in existing §50.59(a)(2)(ii),
renumbered to §50.59(c)(2)(v) in the
proposed rule, be revised to read
``(would) create the possibility for a
design basis accident of a different type
from any previously evaluated in the
final safety analysis report (as
updated).'' This change had two partsÐ
the first, changing from may be created
to ``would create'' and the second being
the insertion of the phrase ``design
basis.'' The purpose of the first change
was to provide some flexibility to
licensees. Thus, rather than having to
prove that an accident had not been
created, under this rule language, a
licensee would need to request a license
amendment only if it could be
reasonably concluded that the
possibility of an accident of a different
type is created by the change, test, or
experiment. The intent of the second
change was to indicate that in referring
to ``accidents'' in §§50.59 and 72.48, the
Commission had in mind creation of
accidents of the likelihood and
significance of those that, had the
possibility already existed, would have
been a design basis accident in the
FSAR. Thus, ``accidents'' that would
require multiple independent failures or
other circumstances in order to ``be
created'' would not fall within this
criterion.
For an accident to be of a different
type, a few commenters thought that the
accident must result in a new or greater
release path than originally considered,
result in a new fission product barrier
failure mode, or create a new sequence
of events that results in significant
cladding failure, ``such that the accident
would have been included if the FSAR
were being written today.'' The
Commission agrees that these are useful
considerations for determining whether
a change results in an accident of a
different type.
One commenter noted that for certain
older facilities, the term ``design basis
accident'' was only applied to a very
small set of events. Other commenters
thought that accidents must be
``credible'' to be ``created.'' Another
commenter was concerned that a
slightly different initiator leading to the
same design basis accident might be
viewed as an accident of a different
type.
One commenter stated that ``accident
of a different type'' should be changed
to ``accident with a different result,'' for
consistency with the criterion on
malfunction. However, the Commission
also notes the similarity with the
criterion in §50.92 (for no significant
hazards consideration determination).
Allowing changes that result in an
accident of a different type (even if the
result has previously been analyzed)
appears inconsistent with the criterion
in §50.92.
The Commission has concluded that
use of the modifier ``design basis'' with
respect to accidents of a different type
in the rule language may be confusing
because, by the terms of the rule,
accidents of a different type are distinct
from those (design basis) accidents
evaluated in the FSAR. Therefore, in the
final rule, the Commission removed the
phrase ``design basis.'' The Commission
agrees that the accident must be credible
in the sense noted above, of having been
created within the range of assumptions
previously considered (e.g., random
single failure, loss of offsite power, no
reliance on non-safety-grade equipment,
etc.), and that a new initiator of the
same accident is not a ``different type''
(but may affect the frequency of that
accident under §50.59(c)(2)(i)).
Therefore, the final rule uses the same
language as is currently contained in the
existing rule, concerning accidents of a
different type, except for changing the
phrase ``possibility ** * may be
created'' to ``would create the
possibility.''
Need for Definition of Accident
In addition, the Commission had
requested comment as to the need for a
definition of accident, and offered a
specific definition for comment. The
term ``accident'' also appears in other
evaluation criteria, specifically,
§§50.59(c)(2)(i) and 50.59(c)(2)(iii), in
the context of accidents previously
evaluated in the FSAR.
Several comments were received on
the proposed definition of accident.
Most commenters felt that a definition
in the rule was not necessary, and most
also disagreed with the specific
definition offered in some respect.
Commenters generally agreed that
accidents include design basis accidents
(typically analyzed in Chapters 6 and 15
of the FSAR), anticipated occupational
occurrences, external events that the
plant is required to withstand and other
special events that are analyzed to
demonstrate safety. Included within the
set of accidents are those scenarios for
which requirements have been
established for the facility either to
withstand or cope with the event.
Notable examples include pressurized
thermal shock events (§50.61),
anticipated transient without scram
(§50.62) and station blackout (§50.63).
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Commenters also noted that external
events, such as earthquakes, high winds,
floods, and missiles can be treated as
causes of malfunctions of SSC, rather
than accidents. Some suggested that
examples or a list of accidents could be
presented in the implementation
guidance.
The Commission concludes that a
definition of accident is not necessary in
the final rule and that examples of
accidents are best discussed in rule
implementation guidance.
I. Possibility of a Malfunction of
Structures, System, or Components
Important to Safety With a Different
Result From Any Previously Evaluated
in the Final Safety Analysis Report (as
Updated) is Created
In the proposed rule, the Commission
modified the remaining part of existing
§50.59(a)(2)(ii), concerning
malfunctions of a different type by
creating a new criterion (vi), that would
require approval if a change, test, or
experiment would ``create a possibility
for a malfunction of equipment
important to safety with a different
result than any evaluated previously in
the final safety analysis report (as
updated).''
Comments were supportive of the
change from ``different type'' to
``different result,'' and of the change
from ``may be'' to ``is'' created. Some
commenters objected to the insertion of
the phrase ``important to safety'' and
suggested other phrases, such as ``safety-
related'' or ``FSAR-described.'' Others
suggested that the terminology in
§§50.59 and 72.48 should be made
consistent (the former refers to
equipment; the latter to systems,
structures or components).
In the final rule, The Commission has
revised the existing criterion to read
``create a possibility for a malfunction of
an SSC important to safety with a
different result from any previously
evaluated in the final safety analysis
report (as updated).'' The Commission
concludes that the term ``SSC'' is
commonly used in both parts 50 and 72
and is well-understood, and that
equipment was an older term that does
not have a unique meaning requiring its
use. The modifier ``important to safety''
was considered as always being part of
the criterion in practice, and that its
omission from the rule was viewed as
editorial and not substantive. Other
terms might have the effect of limiting
or broadening the scope of SSC to be
considered. The Commission notes that
since the overall scope of §50.59 is the
facility as described in the FSAR, there
is no need to use that phrase in
characterizing which SSC need be
considered with respect to
malfunctions.
Guidance for Malfunction With a
Different Result
The proposed rule discussion further
stated that this determination should be
made either at the component level, or
consistent with the failure modes and
effects analyses (FMEA), taking into
account single failure assumptions, and
the level of the change being made.
Several commenters stated that this
guidance should be revised to refer only
to the failure modes and effects analysis
in the FSAR, and not to specify the
component level. The Commission
agrees that this criterion should be
considered with respect to the FMEA,
but also notes that certain changes may
require a new FMEA, which would then
need to be evaluated as to whether the
effects of the malfunctions are
bounding.
J. Replacement Criteria for ``Margin of
Safety as Defined in the Basis for Any
Technical Specification is Reduced''
The phrases ``margin of safety'' and
``as defined in the basis for any
technical specification'' in the third
criterion in existing §50.59(a)(2) have
been the subject of differing
interpretations for a number of years
because §50.59 does not define what
constitutes a margin of safety or a basis
for any technical specification in the
context of §§50.59 and 72.48.
The Commission continues to believe
that changes representing a potentially
significant decrease in certain margins
should require NRC review and
approval prior to their implementation.
Margins within the plant design and in
the established licensing basis exist on
many levels. There are margins from the
assumptions of initial conditions,
conservatisms such as computer
modeling and codes to account for
uncertainties, allowances for instrument
drift and system response time,
redundancy and independence of
components. Margins are built into the
facility to account for routine plant
fluctuations and transients and response
to accident conditions. Margins also
exist in the established regulatory
acceptance criteria to be met for
response to various accidents and
transients. The acceptance criteria are
established at a value that accounts for
uncertainty about physical properties
and other variability. As a result,
substantial margins are provided by the
regulatory envelope within which a
plant has demonstrated its ability to
respond to a spectrum of design basis
accidents. In sum, not every margin is
important to assuring safety such that
changes in that margin must be
reviewed and approved by the NRC
prior to their implementation. However,
the Commission recognizes that
precisely delineating the margins for
which changes would require prior NRC
review and approval is a difficult task.
A change criterion which does not
directly refer to margins, but which
nonetheless indirectly assures that
important design and licensing basis
margins are not changed without prior
NRC review and approval, is an
acceptable alternative that would meet
the Commission's goal of assuring
regulatory review of potentially
significant changes to certain margins.
Such an approach avoids having to
describe in the rule the margins of
regulatory interest, and the nature of the
change in margin for which prior NRC
review and approval would be required.
In the proposed rule, the Commission
solicited public comment on several
options. The Commission also requested
the public to provide alternative means
for control of margin.
Option 1 in Proposed Rule
The first option in the proposed rule
was to control inputs to analyses and
the methods and criteria that establish
TS. Under this option, the Commission
would conclude that the analyses and
information in the FSAR establish the
basis for the margins of safety for the
TS. Thus, the Commission's proposal
would have added a definition for
``reduction in margin of safety
associated with any technical
specification'' and conformed the
criterion for needing a license
amendment in new §50.59(c)(2).
Although this option would maintain
the safety analyses that underlie the TS,
this approach also would have the effect
of giving all input values and
assumptions within the FSAR the
weight of TS (even though they are not
included in the TS), which is
inconsistent with the philosophy in
§50.36. In many instances, changes to
inputs can be accommodated by other
available margins so that the licensing
envelope is preserved. Several
comments expressed strong concern that
this option would be too restrictive, for
the reasons noted above. The
Commission agrees with these concerns
and concludes that the approach is not
consistent with the intent of the original
rule. In this light, this option of
requiring prior NRC approval for any
change to input parameters associated
with TS was rejected as an approach for
the final rule.
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Option 2 in Proposed Rule
The proposed rule contained a second
option that was a proposal to delete the
``margin of safety'' criterion completely.
Instead, the Commission would rely
upon the other criteria in §50.59, as
well as the regulatory requirement that
all changes to TS be reviewed and
approved by the NRC, to assure that
there are no significant adverse changes
to margins in design and operation. If
this option were adopted, the
Commission would argue that there is
no need for prior review of changes that
do not satisfy any of the other
evaluation criteria in view of ``risk-
informed'' insights and greater
understanding of the margins that exist
through meeting the body of regulatory
requirements. The Commission also
sought comment on whether any of the
other evaluation criteria should be
revised if this approach were adopted.
A significant number of comments
were received in support of the proposal
to delete margin of safety as an
evaluation criterion. In support of their
position, commenters noted that TS and
the other six evaluation criteria, in
conjunction with other regulatory
requirements for design, testing, and
operation, make the margin question
moot. The Commission did not adopt
this proposal because of the variability
in existing TS, and uncertainties about
how licensees might gauge the other
evaluation criteria for specific changes.
Option 3 in Proposed Rule
In the Federal Register notice, the
NRC also offered a set of options that
focused on control of margins associated
with results of analyses. Instead of
focusing on the inputs to safety
analyses, these options would focus on
the results of the safety analyses in
order to determine whether changes to
operational characteristics or other
information described in the FSAR (as
updated) would reduce the level of
protection reflected by the results of
safety analyses.
In developing which results would be
governed by this evaluation criterion,
the Commission considered what
aspects of the facility safety are
controlled by other requirements and
thus what other information might a
``margin'' criterion be intended to
capture. As part of the licensing review
for a facility, the NRC established a level
of required performance (which will be
referred to in this discussion as
acceptance criteria) for certain physical
parameters, such as those that define the
integrity of the fission product barriers
(e.g., fuel cladding, reactor coolant
system boundary, and containment).
Satisfying these acceptance criteria
produces a margin of safety to loss of
barrier integrity. The safety analyses
presented in the FSAR (as updated)
demonstrate that the response of the
barriers to the postulated accidents,
transients, and malfunctions meets the
acceptance criteria. Thus, in
constructing the options for comment,
the Commission suggested a more
explicit linkage between when ``margin
of safety'' needed to be preserved to the
response of the fission product barriers
relied upon to provide protection from
uncontrolled release of radioactivity.
In the range of options, the
Commission also suggested that certain
mitigation system capability, as, for
instance engineered safety feature
performance parameters (flow rates,
efficiencies, etc.) also might be
considered with respect to margin, and
asked for comment whether there were
other parameters that should be
explicitly accounted for in any criterion
on ``margin of safety.''
As part of these options, the
Commission also offered different
approaches to how much flexibility
should be allowed, as for instance,
minimal reductions, or use of limits as
the point at which reductions in margin
would be determined. Also, as
discussed later, the Commission asked
in the proposed rule whether changes to
evaluation methods should also be
controlled.
Comment Summary for Option 3: The
Commission received a large number of
comments on the various suboptions
under Option 3 concerning results of
analyses. With respect to the
identification of those parameters to
control, many of the commenters who
supported a ``margin'' concept based
upon limits for results, believed that the
parameters should be limited to those
that directly affect fission product
barriers and for which there are clearly
defined limits. One commenter thought
that a criterion on margin is not needed
for a reactor that was being
decommissioned. Commenters also
thought that mitigation system
performance was best controlled by
other criteria, such as those concerning
malfunction of SSC, or consequences of
accidents. It was also noted that
important characteristics of mitigation
systems are governed by TS. With
respect to parameters that might be used
under part 72, commenters stated that
these should be those with the potential
to increase the likelihood or the amount
of offsite release, specifically, such
things as fuel and cladding temperature,
cask temperature and internal pressure,
and cask stresses.
For the question as to when NRC
approval is needed, comments can be
grouped into two main themes: those
that are supporting the position
currently included in NEI 96±07 related
to acceptance limits as being the point
of departure for reduction in margin,
and those supporting a new proposal
from NEI. No commenters supported
either a ``no reduction in results'' or a
``minimal'' standard, or any type of
graduated approach such as that
discussed earlier for consequences. As
part of its comments on the proposed
rule, the NEI proposed to replace the
existing margin of safety criterion with
one that states that a change requires
prior NRC approval if it would result in
a design basis limit directly related to
integrity of the fuel cladding, the reactor
coolant system boundary, or the
containment boundary being exceeded
or altered. Their proposal is similar in
several respects to the guidance offered
in NEI 96±07, with respect to using
``limits'' as the point at which a
reduction in margin occurs, and in
focusing on parameters for fission
product barriers as being the instances
where there is margin to protect. The
difference is the concept of ``design
basis limits'' as represented in the FSAR
instead of acceptance limits that might
be found in other documents. Further,
NEI suggested that as part of the rule
changes to adopt this criterion, the NRC
should also delete the third criterion in
§50.92, which states that a
determination of ``no significant hazards
consideration'' cannot be made for
amendments that would involve a
significant reduction in a margin of
safety.
Resolution
In SECY±99±054, dated February 22,
1999, the staff presented an alternate
proposal for the margin of safety
criterion. The staff proposal employed a
concept that used the design basis
capability for a SSC as the determinant
for when prior staff review would be
required. As presented in the final
safety analysis report, there is a design
basis (functions and controlling values
of parameters) that determines the
minimum performance requirements for
SSCs. The controlling value for a
parameter is the point at which
confidence in the capability of the
structure, system or component to
perform its intended safety functions
begins to decrease. For many
parameters, requirements have been
established in TS; for others, which are
not directly controlled or measured,
while certain TS requirements may have
been imposed to keep values within
required ranges, inclusion of a criterion
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that verifies that facility changes have
not adversely impacted design basis
capability provides assurance of
completeness beyond the requirements
for approval of TS changes.
The staff was supportive of the NEI
concept of using the design basis as the
determinant of when prior NRC
approval was needed. The staff proposal
was a modification of the suggested NEI
approach that would focus on the
effectiveness of systems to protect
barriers. The staff thought that the rule
language as offered by NEI could be
viewed too narrowly, and might not
ensure that changes affecting
performance of mitigation and support
systems were appropriately evaluated
with respect to their roles in protecting
integrity of the barriers. Therefore, the
staff's proposal was more explicit about
the design basis capabilities of the SSC
being used to determine whether
approval of a change was needed. The
principal difficulty with this proposal
was uniquely identifying the design
basis capabilities for all SSCs that
would need to be satisfied in order to
implement the concept.
Since the time that SECY±99±054 was
submitted to the Commission, the NRC
has gained a greater understanding of
the NEI proposal and how it would be
implemented, and, in particular, how it
would be used to assess changes to
mitigation systems and support systems.
Although the NRC agreed that the
process described in the NEI comment
letter of December 21, 1998, would be
sufficient to ensure that changes to other
systems are appropriately examined
with respect to impact upon the
barriers, it was not apparent that the
specific rule language suggested would
require licensees to implement such a
systematic approach to examination of
design basis limits.
Therefore, the approach contained in
the final rule is a combination of the
NEI proposal contained in its comment
letter and the staff proposal contained in
SECY±99±054. In the final rule, the
Commission is eliminating the existing
criterion on reduction of margin of
safety. In its place, the Commission is
adding a new criterion (vii) that requires
prior NRC review of changes that result
in a design basis limit related to the
integrity of the fission product barriers
being exceeded or altered.
The final rule also contains a new
criterion (viii) related to the use and
control of evaluation methods (see
below). These two criteria together in
place of a criterion on margin of safety
explicitly cover those margins that the
Commission believes are important to
address in this evaluation processÐthe
first being the margin that exists in the
limits that are to be met, and the second
being the margin that exists from the
conservatisms included in the methods
used to demonstrate that requirements
are met. Each of these criteria are
discussed below.
The Commission concludes that the
new criteria (vii) and (viii) together will
maintain safety because they will
preserve the design basis capabilities
that protect the integrity of important
fission product barriers, and thus those
features that protect against release of
radioactive material. The rule will also
control the analyses and assessment
process through control of the methods
and will assure that the required
response of the barriers as previously
established by NRC review will be
maintained.
The Commission does not plan to
make any changes to the criterion in
§50.92(c)(3), which provides that
license amendments involving a
significant reduction in a margin of
safety do not meet the criteria for a ``no
significant hazards consideration''
determination as discussed in section M
below.
Final Rule Language
New Criterion (vii)
New criterion (vii) would require a
prior NRC review of any change that
would ``result in a design basis limit for
a fission product barrier as described in
the FSAR (as updated) being exceeded
or altered.'' For purposes of
implementation of this criterion, the
Commission defines design basis limit
for a fission product barrier as the
controlling numerical value for a
parameter established during the
licensing review as presented in the
final safety analysis report for any
parameter(s) used to determine the
integrity of a barrier. Typically, the
controlling value for the parameter is set
at a point far enough away from failure
that there is confidence in the integrity
of the barrier. As a partial substitute for
the previous ``reduction in margin''
criterion in the former §50.59(a)(2)(iii),
a change which does not exceed or alter
a design basis limit for a fission product
barrier does not involve any reduction
in the margin of safety.
The Commission did not retain the
suggested wording from commenters for
criterion (vii) which might suggest that
the evaluation can be limited to those
changes that are directly related to fuel
cladding, reactor coolant system
boundary, and containment boundary.
The Commission believes that a broader
initial assessment of parameters is
necessary than that which might be
suggested by the term ``directly related.''
All changes that might affect the design
basis limits, including changes to
parameters within mitigation and
support systems, must be evaluated for
their effects upon the design basis limits
for the barriers. Further, the
Commission used the term ``fission
product barrier,'' rather than listing the
specific barriers for operating power
reactors as used by NEI, so that the rule
language would be appropriate for all
Part 50 facilities (including non-power
reactors, and reactors undergoing
decommissioning). The more general
terminology is also appropriate for the
part 72 facilities.
New criterion (vii) narrows the focus
for when prior NRC approval is required
to those changes which result in the
specific limits that relate directly to the
performance of fission product barriers
being exceeded or altered. For power
reactors, these barriers are generally
limited to the fuel cladding, the reactor
coolant system pressure boundary and
containment. For a reactor undergoing
decommissioning, where the fuel is
stored in the spent fuel pool, the barrier
would be the fuel cladding. For non-
power reactors, the fission product
barriers would include, as applicable to
the specific reactor, the fuel cladding,
the reactor tank, and the reactor room,
building, confinement, or containment.
The proposed criterion (vii) is equally
applicable to independent spent fuel
storage facilities or spent fuel storage
cask designs in part 72. The particular
parameters or barriers would be
specified in terms of the barriers against
release of radioactivity afforded by fuel
storage facilities. For instance, these
would include calculated fuel
temperature or cladding oxidation, and
stresses (or pressures) on the cask
structure.
Although the list of fission product
barriers includes containment and other
features that prevent the release of
radiation, the design basis limits for
these barriers are for parameters such as
pressure. The determination of resultant
radiological consequences from leakage
through or breech of these barriers is the
subject of criteria (iii) and (iv), rather
than criterion (vii).
Further, design basis limits for certain
fission product barriers may not be
applicable to particular facilities or
conditions of the facility (such as
permanently shutdown facilities). The
determination as to the need for
evaluation of particular barrier
parameters or limits depends upon the
safety analyses and information
presented in the FSAR (as updated).
The Commission notes that the new
criterion (vii) does not incorporate the
use of a minimal change concept. The
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modification of the criterion to reflect
design basis limits as a point for
evaluating when prior NRC review is
necessary would not permit small
changes beyond the limits without
review.
With respect to changes relating to the
design basis capability of SSCs to
perform their functions in those
circumstances in which the change does
not cause any design basis limits to be
exceeded or altered, the other
evaluation criteria in §50.59 (as well as
other requirements such as TS or ASME
code requirements) provide the
standards for prior NRC approval of
such changes.
The rule language that provides that
a design basis limit may not be altered
provides important and needed
assurance. Changes that involve
alteration of the design basis limit for a
fission product barrier involve such a
fundamental alteration of the facility
design that a change, even in the
conservative direction, should receive
prior NRC review.
Guidance for Implementation
To satisfy new criterion (vii),
licensees must determine the
parameters that would be affected by the
proposed change. The affected
parameters are not limited to the
specific parameters in the system in
which the change is being made or to
parameters that are only directly linked
to the actual fission product barrier.
Rather, the design parameters must
include an assessment of all affected
parameters, including design parameters
of mitigation and support systems. Once
the parameters are identified, the
licensee must establish whether the
parameters have values established in
the FSAR, whether the parameters are
controlling parameters that are reference
bounds for the design, and whether the
parameter has the potential to affect the
performance of the fission product
barrier. If the specific parameter values
are already subject to controls
established by the TS or other rules or
regulation, those requirements shall be
followed.
After a licensee assesses the
information discussed above, it would
need to identify the specific design basis
limits that could be affected for each of
the identified parameters. After the
licensee completes its assessment of the
change against each design basis limit,
if no design basis limit is altered or
exceeded, criterion (vii) is satisfied, and
a licensee may make the change without
prior NRC review.
Examples
The NRC has selected several
examples to illustrate how the new
criterion (vii) would be implemented. In
these examples, it is assumed that NRC
approval is not required because of
other reasons, such as need for a TS
change, section 50.55a requirements etc.
Example 1: A plant FSAR states that
the function of the auxiliary feedwater
system (AFW) is to provide feedwater
flow to the steam generators following
postulated accidents (e.g., main steam
line break, feed line break, small break
loss-of-coolant accident), or when a
reactor trip occurs coincident with a
loss-of-offsite power. The FSAR states
that 700 gallons per minute (gpm) will
be delivered to the steam generators.
The licensee's accident analyses used
700 gpm to assess the acceptability of
the plant to respond to the accidents
and concluded that no safety limits
were challenged if 500 gpm were
supplied. As a result of recent testing of
the AFW system, the licensee
determines that the pumps can no
longer deliver 700 gpm. The licensee
determines that the AFW pumps can
deliver only 500 gpm at the required
pressure and temperature. The licensee
performs the necessary safety analyses
and confirms that 500 gpm is sufficient
to meet all necessary functions and that
no safety limits would be challenged as
a result of the flow reduction. The
licensee decides to leave the pumps in
the plant as is rather than replace the
pumps to restore the originally stated
capability. The licensee revises the
FSAR to state that the AFW system will
deliver 500 gpm during postulated
accidents or for transients involving a
loss-of-offsite power.
Under the new criterion (vii), the
licensee would have to assess the
impact of the reduced flow rate on the
design limits of the fission product
barriers. The licensee would have to
identify the system parameters that
would vary as a result of the changes in
AFW system performance, identify the
specific design limits that have the
potential to affect the fission product
barrier performance, and complete the
analyses to determine whether the
specific design limits for the fission
product barriers would be challenged.
In this example, it is assumed that the
licensee did not change the method of
evaluation for the safety analyses. If the
licensee had used a different
methodology from that used initially in
establishing that the limits were met,
then, the licensee may have to submit
the revised analyses under criterion
(viii) of the revised rule.
For this example, the licensee would
have to complete the evaluations
required by §50.59 but would not have
to submit a license amendment request
to lower the expected flow rate of the
AFW system, from that stated in the
FSAR, to the lower as-found value, nor
would a licensee have to request an
amendment to remove the old pumps
and replace the pumps with new pumps
that provide the lower capacity assumed
in this example. The basis for this
conclusion is that the licensee analyses
determined that the design limits of the
fission product barriers would not be
challenged and, therefore, that the
fundamental basis for the staff's initial
safety conclusion is maintained.
Example 2: A facility FSAR states that
some of the functions of the component
cooling water system are to provide
cooling water flow to the reactor coolant
pump seals and to the shell side of the
residual heat removal system (RHR) heat
exchangers. The FSAR states that the
CCW system provides 400 gallons per
minute, 100 gpm for the seals and 300
gpm for the RHR heat exchanger. The
licensee has recently obtained a new
reactor coolant pump seal which
requires an additional 25 gpm of cooling
flow. The licensee plans to revise the
flow distribution such that 125 gpm is
directed to the seals, and 275 gpm to the
RHR heat exchangers. The licensee
performs analyses to determine that
with the reduced CCW flow to the RHR
heat exchangers, the RHR system can
still perform its required functions with
required limits, as for example,
removing sufficient decay heat to cool
down within required time frames,
keeping post-accident temperatures
within required limits, etc. The licensee
would satisfy criterion (vii) and be able
to make this change under §50.59.
Example 3: A licensee discovers an
error in the primary system pressure
boundary piping fatigue calculation
performed to demonstrate compliance
with the ASME Code requirements. A
corrected calculation shows that the
fatigue criterion would be exceeded (for
the postulated FSAR events). A change
to the licensing basis to accept revised
fatigue criteria would require review
under criterion (vii) because the design
basis limit for one of the fission product
barriers (reactor coolant system piping)
would be exceeded or altered. (This
change would also not satisfy criterion
(i), ``minimal increase in frequency of
occurrence of an accident'' because of
potential failure of piping due to fatigue
cracking, leading to loss of piping
system integrity.)
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New Criterion (viii)ÐControl of
Evaluation Methods
In the proposed rule notice as part of
the options presented on margin of
safety, the Commission had discussed
the issue of controlling methods (also,
as noted, the proposed rule had
explicitly stated that changes to
methods were changes to the facility,
and as such, required §50.59
evaluations). Specifically, the
Commission sought comment on
whether the rule should include a
statement that ``all analyses and
evaluations for assessing the impact of
plant changes must be performed using
methodology and analytical techniques
which are either reviewed and approved
by the NRC or which are shown to meet
applicable review guidance and
standards for such analyses.''
Five commenters stated that methods
should not be controlled by §50.59
because the limits (e.g., acceptance
limits) are conservative. These
commenters thought that licensees
should be allowed to use methods that
are accepted by the NRC Standard
Review Plan or other processes, without
the need for prior NRC approval. A few
commenters agreed that methods should
either be reviewed and approved by
NRC (or meet applicable standards);
produce results that are consistent with
the licensing basis methods; or that
changes to methods should be reviewed
as separate changes under §50.59.
The Commission concludes that
control of methods is essential in
assuring a consistent application of the
change review process, especially in
light of the flexibility being provided by
changes to the other evaluation criteria,
such as having criterion (vii) that uses
design basis limits being exceeded as
the point at which NRC review is
required instead of the ``margin of
safety'' criterion. Although the
Commission agreed that changes to
methods should be reviewed as separate
changes, the other evaluation criteria do
not provide a standard that could be
used to determine when changes to
methods should be reviewed by NRC.
While the NEI proposal would have
controlled the methodologies through
regulatory guidance, the Commission
did not judge that process to provide
sufficient rigor to assure uniform
implementation of the requirement. A
statement that the analysis should meet
applicable standards was considered,
but was ultimately rejected as being too
vague. Therefore, the Commission has
added criterion (viii) to be specifically
used for changes to methods of
evaluation.
Final Rule Language
New criterion (viii) will require prior
NRC review of any change in a
methodology or evaluation method that
``results in a departure from a method of
evaluation described in the FSAR (as
updated) used in establishing the design
bases or in the safety analyses.''
Definitions and Guidance
For the purposes of this rule, a
departure from a method of evaluation
described in the FSAR (as updated)
used in establishing the design bases or
in the safety analyses means (1)
changing any of the elements of the
method described in the FSAR (as
updated) unless the results of the
analysis are conservative or essentially
the same; or (2) changing from a method
described in the FSAR to another
method unless that method has been
approved by NRC for the intended
application. Results from a changed
method are conservative relative to
results from the previous method, if
closer to the limits or values that must
be satisfied to meet the design bases.
Results are ``essentially the same'' if
they are within the margin of error
needed for the type of analysis being
performed, even if tending in the non-
conservative direction. Results are
essentially the same if the variation in
results because of the change to the
method is explainable as routine
analysis sensitivities, and the
differences in the results are not a factor
in determining whether any limits or
criteria are satisfied. The determination
can be made through benchmarking
(new vs. old method), or may be
apparent from the nature of the changes
between the methods. When
benchmarking a method to determine
how it compares to the previous one,
the analyses that are done must be for
the same set of plant conditions,
otherwise, the results may not be
comparable. Approval for intended
application includes assuring that the
approved method was approved for the
type of analysis being conducted,
generically approved for the type of
facility using it, and that all terms and
conditions for use of the method are
satisfied.
The rule words were chosen to allow
licensees only a small degree of
flexibility in methods where the results
are tending in the non-conservative
direction, without burdening either the
licensee or the NRC with the need to
review very small changes that are not
important with respect to the
demonstrations of performance that the
analyses are providing. The intent is to
limit the need for review to those
changes to methods that could impact
upon the acceptability of performance
were the results to be at the limiting
values.
By limiting the methods to those
described in the FSAR, and to those
used for design bases and safety
analyses, the Commission concludes
that the burden of requiring review is
justified in view of the relaxations in the
other evaluation criteria. Unless the
methods are used in FSAR safety
analyses, as demonstrating that the
facility performance continues to meet
requirements, or to verify conformance
with the design bases, they would not
meet the rule requirements for approval.
Thus, for example, if a licensee chose to
perform sensitivity studies, or to
examine alternative approaches for a
change being contemplated, or included
other analyses in the FSAR for reference
purposes, these methods would not be
subject to the rule. It is at the point in
time that the revised method becomes
the means used for purposes of
satisfying FSAR safety analysis or
design bases requirements that the
approval (if the noted conditions are not
met) would become necessary.
The Commission has included a
definition of ``departure'' in the
definitions section of the rule such that
the intended meaning for purposes of
§50.59 is clearly understood.
Design bases as used in criterion (viii)
is that information meeting the
definition contained in 10 CFR 50.2,
and in particular, those controlling
values that are restraints derived from
generally accepted practices for
achieving functional goals, or
requirements derived from analysis of
the effects of a postulated accident for
which a SSC must meet its functional
goals. Safety analyses are those
evaluations that demonstrate that
acceptance criteria for the facility's
capability to withstand or to respond to
postulated events are met.
Thus, this criterion applies to those
methods of evaluation used for
demonstrating that design basis limits
for fission product barriers are met, for
other analyses such as radiological
consequences that are part of the safety
analyses, and for analyses that
demonstrate that functional goals for
SSC are met. These would include those
analyses that show that SSC will
function under limiting conditions such
as natural phenomena, environmental
conditions, dynamic effects, and so
forth. However, as noted in the rule
language, only those methods that are
used in establishing the design bases or
in the safety analyses fall within the
criterion. In addition, the Commission
notes that changes to time-limited aging
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analyses and evaluations of aging
management programs required by
§§54.21(d) and 54.37(b), require
evaluation with respect to criterion (viii)
to the extent that evaluation methods for
these analyses are described in the
FSAR supplement.
To assure consistent implementation
of criterion (viii), the Commission
believes that it is important to clearly
distinguish between methods of
evaluation and input parameters to the
methods. Methods of evaluation means
the calculational framework for
evaluating behavior or response of the
reactor or any SSC. This includes the
following (to the extent that they are
described or applicable for a particular
method):
ÐData correlations
ÐMeans of data reduction
ÐPhysical constants or coefficients
ÐMathematical models
ÐSpecific assumptions in a computer
program
ÐSpecified factors to account for
uncertainty in measurements or data
ÐStatistical treatment of results
ÐDose conversion factors and assumed
source term(s)
Input parameters are defined as those
values derived directly from the
physical characteristics of structures,
systems or components, or processes in
the plant. These would include such
things as: Flow rates, temperatures,
pressures, dimensions or measurements
(e.g., volume, weight, size), or system
response times. Changes to input
parameters (that are described in the
FSAR) are to be evaluated as facility
changes, and criterion (viii) would not
be applicable. Additional guidance will
be provided in the implementation
guidance to describe the specific
elements of the evaluation methods or
methodology that would require review
and to clearly define specific types of
input parameters. The NRC intends to
work closely with stakeholders to revise
the existing guidance related to
implementation of §50.59 to reflect
these definitions.
The rule requirements for evaluation
methods would allow for use of generic
topical reports as not being a
``departure,'' provided that the topical
report is applicable to the facility, and
is used within the terms and conditions
specified in the approved topical report.
The Commission believes that with
the guidance concerning ``evaluation
methods'' and the definition of
departure, licensees have the capability
to perform analyses as needed without
being unduly burdened by the need for
NRC review, while still preserving those
inherent conservatisms in the methods
that provide the confidence that safety
is maintained when the parameters are
calculated to be at their design basis
limits and that SSC capability continues
to meet design basis requirements.
Examples
Example 1: The FSAR states that a
damping value of 0.5 percent is used in
the seismic analysis of safety-related
piping. The licensee wishes to change
this value to 2 percent to reanalyze the
seismic loads for the piping. Using a
higher damping value to represent the
response of the piping to the
acceleration from the postulated
earthquake in the analysis would result
in lower calculated stresses because the
increased damping reduces the loads.
Since this analysis was used in
establishing the seismic design bases for
the piping, and since this is a change to
an element of the method that is not
conservative and is not essentially the
same, the NRC concludes that this
change would require approval under
criterion (viii). On the other hand, had
NRC approved an alternate method of
seismic analysis that allowed 2 percent
damping provided certain other
assumptions were made, and the
licensee used the complete set of
assumptions to perform its analysis,
then the use of the 2 percent damping
under these circumstances would not be
a departure, under the second part of
the definition.
Example 2: The licensee wishes to use
an inelastic analysis procedure, not
previously used in its seismic analyses
as described in the FSAR, to
demonstrate that the structural
acceptance criteria are met for cable
trays. NRC concludes that this would be
a departure from the methods of
evaluation and that it would not be
essentially the same because the revised
analysis would predict greater capacity
than would the previous analysis.
Therefore, this change would require
NRC approval.
Example 3: The licensee wishes to
change a non-LOCA FSAR Chapter 15
transient methodology. The
methodology is being changed to a
different vendor's NRC approved
method. The new vendor's method has
been approved generically for the
particular reactor type (e.g., 2 loop
PWR) and for the particular transient
being analyzed. The analysis is being
performed in accordance with all the
applicable limitations and restrictions.
The licensee can make this change
without prior NRC approval because
using a generically approved method for
the purpose it was approved, while
meeting all the limitations and
restrictions, is not a ``departure.''
Subsequent plant changes can then be
evaluated using this new method and
the other seven criteria in §50.59.
Example 4: The licensee wishes to
change an analysis described in the
FSAR which states that adequate net
positive suction head (NPSH) is verified
by analysis without crediting
containment overpressure. The new
analysis will assume that five pounds of
overpressure is credited in calculation
of available NPSH. The revised analysis
predicts more (five additional pounds
of) available NPSH for the pumps, a
result further from the limit (the
required NPSH) for an analysis that
establishes part of the design bases for
the pumps as being capable of
performing their required function
under the range of expected conditions.
This change can not be made without
prior NRC approval because a change in
an element of a method described in the
FSAR, used to establish the design
basis, that is not conservative, or
essentially the same, is a ``departure.''
Example 5: The licensee wishes to
change an evaluation method described
or incorporated by reference in the
FSAR Chapter 15 transient analysis. In
an attempt to remove some of the
conservatism associated with the
analysis, the change the licensee is
contemplating is removal from the
analysis of consideration of certain
instrument uncertainties for a few
parameters, by assuming nominal values
instead. By not accounting for the
greater range of the parameter
(including the uncertainties), the
analysis predicts response further from
the limit to be satisfied. The treatment
of uncertainties was an element of the
method described in the FSAR, and,
therefore, this change can not be made
without prior NRC approval because a
change in an element of a method
described in the FSAR, used in the
safety analysis, that is not essentially
the same is a ``departure.''
On the other hand, if an instrument in
the plant were replaced with a different
one, the assumed uncertainty in the
analysis for that instrument could be
used in the analysis without prior NRC
review, using the other seven §50.59
criteria rather than criterion (viii),
because this is an input change rather
than a model change. How the
uncertainties are treated in the analysis
is part of the method. The range of
values of the uncertainties associated
with particular instruments is a
characteristic of the facility and is thus
an input parameter.
K. Safety Evaluation
The Commission proposed to delete
the word ``safety'' in referring to the
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1Effects of changes includes appropriate
revisions of descriptions in the FSAR such that the
FSAR (as updated) is complete and accurate.
required evaluation for determining
whether the change, test, or experiment
requires a license amendment. A similar
change was proposed for §50.71(e),
which presently refers to safety
evaluations either in support of license
amendments or of conclusions that
changes did not involve USQs.
The Commission also proposed to
change ``safety evaluation in support of
license amendments'' to ``safety analysis
in support of license amendments.'' The
second part of the existing phrase would
be revised to refer to the ``evaluation
that changes did not require a license
amendment in accordance with
§50.59(c)(2) of this part.'' Conforming
changes in Part 72 to revise the language
to refer to ``evaluation'' were also
proposed.
Commenters were generally
supportive of these proposed changes. A
few noted that as with the term ``USQ,''
a simple process should be adopted for
revision of TS that use the term safety
evaluation (this issue is discussed under
Section A(4)). Other clarifying wording
changes were included as a result of the
comments, as for instance, referring to
``approved'' license amendments rather
than to ``requested'' license
amendments to make clear that the
updates, as well as subsequent §50.59
evaluations, should be based upon what
has been approved (and implemented),
not on what a licensee may have
proposed for approval, but that has not
been approved.
The final rule includes these changes
offered in the proposed rule for
§50.71(e); in addition, the term
``approved'' was used in reference to
license amendments. The final rule
language for §50.71(e) is presented in
Section L, which also discusses other
aspects of the requirements for FSAR
updating.
L. Reporting and Recordkeeping
Requirements
Records
Requirements for records for
evaluations performed under §50.59,
and for submittal of a summary report
are being moved to paragraph (d) as part
of this rulemaking. In the final rule, the
Commission has simplified the rule text
concerning records. Although the text is
simpler, there is no change in which
records are being required. That is, the
Commission views the phrase ``made
pursuant to paragraph (c)'' as referring
to those changes, tests, and experiments
that require evaluation against the
criteria (for example, because they
involve the facility as described in the
FSAR), but not to those other activities
or changes that are determined to not
fall within these required evaluations
(as for instance, being screened out). As
noted in Section K above, the rule now
refers to ``evaluations'' not to ``safety
evaluations.''
In addition, the Commission had
proposed a change to the record
retention requirements in existing
paragraph §50.59(b)(3) (renumbered by
this rulemaking to (d)(3)). The change
would add to the requirement that the
records of changes to the facility be
maintained until the termination of the
license, the following statement ``or
until the termination of a license issued
pursuant to 10 CFR part 54, whichever
is later.'' Commenters were supportive
of this proposal, and the final rule
section is unchanged from the proposed
rule in this regard.
Summary Report
Simplified text was also included in
§50.59(d)(2), concerning submittal of
the summary report. The existing text
required submittal annually, or along
with the FSAR update (which could be
up to 24 months between submittals), or
at such other frequencies as specified in
the license. The Commission sees no
need for such variability in submittal
dates, and believes that a 24 month
interval is acceptable for submittal of
the summary report. Licensees may
submit reports more often if they wish.
If a licensee has a shorter time specified
in its license, that licensee may request
that the requirement be removed so that
the rule frequency would be applicable.
The 24 month frequency is also
included in the part 72 sections, as
requested by several commenters.
Updates to the Final Safety Analysis
Report
In the proposed rule, the Commission
proposed to supplement the reporting
requirements in §50.71(e) on ``effects''
of changes to require that in the FSAR
update submittal (with the replacement
pages), the licensee shall include a
description of each change affecting that
part of the SAR that provides sufficient
information to document the effect of
the change upon the probability or
consequences of accidents or
malfunctions, or reductions in margin
associated with that part of the SAR.
The reason for this proposal was that
the Commission was concerned about
the potential cumulative effect of
minimal increases. Since some increases
are allowed in probability and
consequences, the Commission thought
that these rule changes would place
greater importance on: (1) Complete and
accurate SAR updating; (2) the
licensee's evaluation process taking into
account other changes made since last
update; (3) the licensee's screening
process examining plant changes to
determine whether they are indeed
changes requiring evaluation; and (4)
reporting requirements so that staff can
assess the ongoing nature of cumulative
impact.
The issue discussed in the proposed
rule was how the NRC could best
oversee the process such that several
``minimal'' changes do not result in
unacceptable results. In the proposed
rule, the Commission proposed
requiring licensees to report effects of
changes in the FSAR update submittal
in accordance with §50.71(e) in a
different manner to facilitate evaluation
of cumulative effect.
A large number of commenters stated
that this proposal was burdensome and
unnecessary in view of the minimal
standards. Further, commenters thought
that this provision would require them
to perform additional evaluations of the
cumulative effects, or to numerically
gauge the result of increases to
probability that were judged on a
qualitative basis. Others stated that
when analyses were performed, such as
for consequences or performance of SSC
against limits, the existing update
requirements would specify that the
effects of these analyses be included in
the update. The Commission agrees that
the burden associated with the proposed
rule change is not warranted in view of
the specific criteria adopted and the
existing update requirements. Therefore,
the final rule does not contain such
language.
Other wording changes for §50.71(e)
were discussed under section K.
Therefore, the following language is in
the final rule for this section:
(e) Each person licensed to operate a
nuclear power reactor pursuant to the
provisions of §50.21 or §50.22 of this part
shall update periodically, as provided in
paragraphs (e)(3) and (4) of this section, the
final safety analysis report (FSAR) originally
submitted as part of the application for the
operating license, to assure that the
information included in the FSAR (as
updated) contains the latest information
developed. This submittal shall contain all
the changes necessary to reflect information
and analyses submitted to the Commission
by the licensee or prepared by the licensee
pursuant to Commission requirement since
the last submittal of the original FSAR, or as
appropriate the last update to the FSAR
under this section. The submittal shall
include the effects1 of: all changes made in
the facility or procedures as described in the
FSAR; all safety analyses and evaluations
performed by the licensee either in support
of approved license amendments, or in
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support of conclusions that changes did not
require a license amendment in accordance
with §50.59(c)(2) of this part; and all
analyses of new safety issues performed by
or on behalf of the licensee at Commission
request. The updated information shall be
appropriately located within the update to
the FSAR.
M. No Significant Hazards
Consideration Determinations
Under §189.a(2)(A), the Commission
may issue and make immediately
effective an amendment to an operating
license if the Commission has made a
determination that the amendment
involves a ``no significant hazards
consideration'' (NSHC), despite the
pendancy of a request for a hearing or
the completion of such a hearing. The
Commission's criteria for determining
whether an amendment involves a
NSHC, as set forth in §50.92(c), are
similar to the current USQ criteria in
§50.59:
(c) The Commission may make a final
determination * * * that a proposed
amendment to an operating license ** *
involves no significant hazards
consideration, if operation of the facility in
accordance with the proposed amendment
would not:
(1) Involve a significant increase in the
probability or consequences of an accident
previously evaluated; or
(2) Create the possibility of a new or
different kind of accident from any accident
previously considered; or
(3) Involve a significant reduction in a
margin of safety.
The Commission has evaluated
whether the NSHC criteria in §50.92(c)
must be modified if the existing criteria
in §50.59 are altered, deleted or
supplanted. The AEA does not define
NSHC, nor does any provision of the
AEA conceptually link the NSHC
concept to any particular standard or
concept. A review of the legislative
history of the ``Sholly amendment''
which modified Section 189.a did not
disclose any reference to §50.59 or a
discussion which links the NSHC
concept and the §50.59 criteria. H.R.
Conf. Rep. No. 97±884, 97th Cong., 2d
Sess. (1982), Sen. Rep. No. 97±113, 97th
Cong., 2d Sess. (1981), H. Rep. No. 97±
22, Part 2, 97th Cong., 2d. Sess. (1981).
The Commission has also evaluated
whether changes to the NSHC criteria to
conform more closely to the revised
§50.59 would facilitate implementation
of the revisions to §50.59, even if
changes to the NSHC criteria are not
required by the AEA. There are three
areas where the current NSHC criteria
diverge from the revised §50.59 criteria:
(i) The current NSHC criteria do not
include the ``malfunction of
components'' criterion in the revised
§50.59; (ii) the NSHC criteria retains a
``significant reduction in margin of
safety'' criterion, which is no longer part
of the revised §50.59; and (iii) the
NSHC criteria do not include the
revised §50.59 criteria (vii) and (viii)
concerning changes to fission barrier
design basis limits, and changes to and
departures from evaluation methods.
Although there may be some conceptual
tidiness in utilizing the same evaluation
factors for changes under §50.59 and
NSHC determinations under §50.92,
nothing in the AEA or the legislative
history requires that the criteria be
identical. Furthermore, the Commission
notes that §50.59 and NSHC address
issues which are fundamentally
different in purpose. Section 50.59 is
focused upon the NRC's regulatory
needs with respect to its review and
approval of licensee-initiated changes,
tests and experiments. By contrast, the
NSHC determination is directed at
determining what license amendments
will require the Congressionally-
mandated 30-day notice in the Federal
Register and completion of any hearing
granted pursuant to the Congressionally-
mandated opportunity for hearing in
Section 189.a. In the Commission's
view, the existing NSHC criteria have
been demonstrated through years of
application to provide a workable
standard for determining the potential
safety significance of a proposed
amendment for the purposes of
determining whether issuance of a
license amendment must await notice in
the Federal Register and completion of
any requested hearing. On balance, the
Commission believes that no changes to
the existing NSHC criteria are necessary
in order to implement the revised
change criteria in the revised §50.59.
Recognizing the difference between
the two sections, the Commission notes
that if a change does not require a
license amendment by virtue of the new
§50.59(c)(2)((vii) and (viii) criteria, then
the change cannot be regarded as
involving a ``significant reduction in a
margin of safety'' under §50.92(c)(3). If
a change does require a license
amendment by virtue of either
§50.59(c)(2)((vii) or (viii), the NRC
would be required to determine whether
the design basis limit for a fission
product barrier being exceeded or
altered, or the departure from the
method of evaluation used in
establishing the design bases or safety
analyses, constitutes a significant
reduction in a margin of safety. With
respect to new §50.59(c)(2)(ii) and (iv),
the Commission regards these criteria as
a substitute for and refinement of the
``malfunction of equipment'' aspect of
the existing §50.59(a)(2)(ii) criterion, for
which there is no parallel provision in
§50.92(c)(2). Therefore, the NSHC
evaluation for license amendments
necessitated by the new §50.59(c)(2)(ii)
and (iv) criteria will be largely the same
as the current process for evaluating
license amendments necessitated by the
``malfunction of equipment'' provision
in the existing §50.59(a)(2)(ii).
N. Part 52 Changes
In the proposed rule, the Commission
had proposed to revise appendices A
and B to part 52 to conform with the
proposed changes to §50.59 concerning
the evaluation criteria for when prior
NRC approval is required for changes to
certain Tier 2 information in plant-
specific design control documents.
Two commenters believe that the
changes to part 52 needed to be
expanded to either include certain
provisions or definitions, or to refer to
§50.59 to incorporate them. The
Commission has decided to defer
consideration of the changes in the
proposed rule for part 52. The
Commission anticipates other rule
changes for Part 52 arising from an
ongoing lessons-learned review.
Further, the proposed design
certification rule for the AP600 design
being issued for public comment will
emulate the two design certification
rules in appendices A and B.
Accordingly, the Commission will
consider these proposed changes in an
integrated manner later.
O.1. Part 72 Changes
This section first discusses the
changes offered in the proposed rule on
part 72, then discusses the comments
received and the resolution and final
rule language. The comments and rule
language are discussed under
subheadings relating to the specific
requirements, such as for evaluation of
changes, FSAR updating, and other
conforming changes. A discussion of
petition for rulemaking (PRM 72±3),
submitted by Ms. Fawn Shillinglaw, and
how it relates to the changes to part 72
is contained in section O.2.
Changes Presented in the Proposed Rule
For part 72, in the proposed rule, the
Commission proposed changes to
§72.48 conforming with those made to
§50.59 and proposed to expand the
scope of §72.48 so that holders of a
Certificate of Compliance (CoC)
approving a spent fuel storage cask
design also would be subject to the
requirements of this section. The
Commission envisioned that a general
licensee who wants to adopt a change to
the design of a spent fuel storage cask
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it possessesÐwhich change was
previously made to the generic design
by the certificate holder under the
provisions of §72.48Ðwould be
required to perform a separate
evaluation under the provisions of
§72.48 to determine the suitability of
the change for itself.
Certificate holders would be required
to keep records of such changes as are
allowed under §72.48. New reporting
requirements for certificate holders
would be added in §§72.244 and
72.248, similar to existing requirements
imposed on licensees in §§72.56 and
72.70, respectively.
In addition to these changes to
§72.48, the Commission proposed
making changes in other sections of part
72 as follows:
In §72.3 the definition for
independent spent fuel storage
installation (ISFSI) would be revised to
remove the tests for evaluation of the
acceptability of sharing common
utilities and services between the ISFSI
and other facilities; and the existing
requirement in §72.24(a) revised to
reference shared common utilities and
services in the applicant's assessment of
potential interactions between the ISFSI
and another facility. Proposed changes
to §72.56 would be conforming changes
to those made to §50.90. Changes to
§§72.9 and 72.86 are conforming
changes due to the proposed addition of
new §§72.244, 72.246, and 72.248. The
change to §72.212(b)(4) would be a
conforming change necessitated directly
by the change to §50.59, as this section
in part 72 refers to §50.59 with respect
to evaluations for the reactor facility at
which site the ISFSI is located.
In the proposed rule, §72.70 was
proposed for revision to conform to
§50.71(e). Requirements would be
added on standards for submitting
revised Final Safety Analysis Report
(FSAR) pages. Requirements would also
be established for reporting changes to
procedures. New reporting requirements
for certificate holders would be added
in §§72.244 and 72.248, similar to
existing requirements imposed on
licensees in §§72.56 and 72.70,
respectively.
New §§72.244 and 72.246 would be
added to subpart L, to provide
regulations on applying for, and
approving, amendments to CoCs. A new
§72.248 would also be added to provide
regulations for the certificate holder on
submitting and updating the FSAR,
which would document the changes it
made to procedures or SSC under the
provisions of §72.48. The new
§72.248(c) would also require, in part,
that updates to the FSAR use revision
numbers, change bars, and a list of
current pages.
Resolution of Comments Received: Of
the 60 comment letters, 10 raised issues
related to part 72. The following is a
summary of those comments and the
Commission's responses:
1. Overall Changes to Part 72
All ten of the commenters were
generally supportive of the changes to
part 72 and the expansion of scope of
§72.48 to include part 72 certificate
holders. Nevertheless, the commenters
indicated that the regulations in part 72
were more restrictive than similar
regulations in part 50. The commenters
pointed to certain part 72 requirements
(i.e., release limits, §72.48 evaluation
criteria on occupational exposure and
environmental impact, and update
frequency and content for §72.48
evaluations and FSAR changes) that do
not exist in part 50 or that are more
stringent than similar part 50
regulations. Overall, the commenters
believe the risk from spent fuel storage
casks and facilities is much less than
from reactors. The commenters
generally recommended that §§72.48
and 72.70 should be more consistent
with §§50.59 and 50.71(e).
The Commission agrees that where
possible the language used in the
respective sections in parts 50 and 72
should be similar. Therefore, except
where unique requirements exist (e.g.,
because §72.48 involves both licensees
and certificate holders, as well as
facilities and spent fuel storage cask
designs, and §50.59 only involves
licensees and facilities), the final rule
has used consistent language in both
parts 50 and 72. The NRC also notes that
the comments on revising the release
limits for part 72 are clearly beyond the
scope of the proposed rule and no
further response is made.
2. §72.48 (Changes, Tests, and
Experiments)
The ten commenters suggested that
the tests in §72.48 should be same as
are used in §50.59; in particular, five
commenters said that the significant
increase in occupational exposure and
significant unreviewed environmental
impact tests were unnecessary and
therefore should be removed. One
commenter indicated the unreviewed
environmental impact test should be
retained, but only for specific licensees.
The Commission agrees that the
occupational exposure test is
unnecessary because licensees are
currently required by §20.1101(b) to
take actions to maintain occupational
exposure as low as is reasonably
achievable. The Commission also agrees
that the significant unreviewed
environmental impact test is
unnecessary. As stated in the Finding of
No Significant Environmental Impact
for this rule, the changes being made in
§72.48 will allow only minimal
increases in probability or consequences
of accidents (still satisfying regulatory
limits) without prior NRC review.
Further, changes which result in more
than minimal increases in radiological
consequences will continue to require
prior NRC approval, including NRC
consideration of potential impact on the
environment. Therefore, consistent with
§50.59, there is no need for this
criterion to be included with respect to
consideration of a change under §72.48
and it has been deleted from the final
rule.
One commenter suggested that the
scope of §72.48 should be limited to
only ``important to safety'' structures,
systems, and components (SSCs), not all
SSCs described in the FSAR. One
commenter suggested the §50.59 term
``equipment important to safety'' should
be used rather than ``SSC important to
safety.'' One commenter suggested the
term ``evaluations'' should be removed
from the definition of the facility in
proposed paragraph §72.48(a)(3)(iii).
The Commission disagrees with these
comments. The term SSCs provides a
better description than equipment and
is consistent with other regulations in
both parts 50 and 72 (as noted earlier,
the Commission is revising §50.59 to
refer to SSC instead of to equipment).
The scope of these §72.48 evaluations
should include all SSCs described in the
FSAR, not just those that are important
to safety. The current regulations in
§72.48 require a scope that includes all
structures, systems, and components
described in the FSAR not just those
``important to safety.'' The Commission
continues to believe that this approach
is necessary to insure that changes to
SSCs considered ``not important to
safety'' do not have a negative impact on
SSCs considered important to safety due
to interactions and interfaces, and do
not cause any adverse impact on public
health and safety. The term ``evaluations
and methods of evaluation'' is necessary
for the reasons previously discussed for
§50.59 changes, and is retained in final
§72.48(a)(2)(iii).
One commenter stated that the term
FSAR should not be used because Part
72 is a one step licensing process and
using the term implies a second review
step is required by staff. The same
commenter added that the discussion of
the FSAR (in the rule) could also imply
that the §72.48 process is not required
to address changes until the licensee has
an FSAR. (The commenter thought the
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proposed rule language suggested that
§72.48 would not apply until after the
FSAR was submitted). Two commenters
identified concerns with the current
requirement for a specific licensee to
update its SAR every 6 months and its
role as a hold point (requiring staff
review) and the requirement to update
the SAR 90 days prior to loading fuel.
Two other commenters suggested that
the order of §§72.48 (a)(2) and (a)(3)
should be reversed and that the term
``required to be included'' should be
deleted from proposed paragraph
(a)(3)(iii).
The Commission has revised §§72.48,
72.70 and 72.248 in response to these
comments. These changes have clarified
the use of the term FSAR to avoid the
interpretation that multiple staff reviews
of this document will be required. The
FSAR being submitted 90 days after
license issuance precludes both a hold
point and an additional staff review.
Further the Commission agrees that
providing a periodic FSAR update every
6 months and a final one 90 days prior
to fuel load was an unnecessary burden,
which does not exist in §50.71(e), and
these requirements have been
eliminated. The Commission agrees that
language was needed to indicate that the
facility or design can be changed using
the new process in §72.48 after a
license is issued and prior to issuing the
FSAR and that has been reflected in the
final rule. Sections 72.48 a(2) and a(3)
have been reversed in order and the
phrase ``required to be included'' has
been deleted for clarity and for
consistency with §50.59.
Several commenters suggested that a
different approach be taken on the
margin of safety; that the terms
``minimal'', ``more than minimal'' or
``significant'' required further
clarification and should be consistent
with §50.59; suggested reports of
§72.48 changes, tests, and experiments
be submitted every 24 months: and that
an implementation schedule be
provided for the final rule.
The NRC agrees that §§50.59 and
72.48 should be as consistent as
possible. Therefore §72.48 has used the
language adopted in response to
comments on §50.59 (see comments on
§50.59 on the use of minimal and
margin of safety terminology). The NRC
agrees that a 24 month reporting
frequency is appropriate. The NRC has
also provided direction in implementing
the final rules.
One commenter suggested that
licensees and certificate holders should
inform each other of changes
implemented under §72.48 that affect a
particular cask design, through the
summary reports rather than through
the FSAR update, as was stated in the
proposed rule. One commenter also
suggested that guidance on the
timeliness of the review to be performed
upon receipt of such changes be
provided.
The NRC agrees with both comments
and has added §72.48 (d)(6)(i)Ð(iii) on
providing copies of §72.48 evaluations
to other interested persons who use the
particular cask design within 60-days of
implementing the change (the proposed
language in §§72.216 and 72.248 on this
point has been deleted). Guidance on
the timeliness of the reviews will be
provided by the NRC along with other
guidance information for §§50.59 and
72.48.
General licensees who have evaluated
a proposed change under §72.48 and
concluded that a CoC amendment is
required, must request that the
certificate holder submit the application
for amendment under §72.244.
Clarifying language was included in
§72.48 on this point.
As a result of other changes made
earlier in §72.48, the section on
recordkeeping was reformatted to
include subsection numbering. As part
of this revision, the text in paragraphs
(d)(3)(i) and (d)(3)(ii) was clarified to
acknowledge those situations where the
facility is no longer being used, but for
which the license has not yet been
terminated.
3. §§72.70, 72.216, and 72.248 (FSAR
Updating)
Several commenters suggested that
the language in §§72.70, 72.216, and
72.248 on updating the FSAR conform
to the language in §50.71(e). Specific
changes requested included requiring a
24-month reporting period, adding a 6-
month cutoff for reporting changes,
clarifying requirements for the initial
submittal of the FSAR, and how no
changes to the FSAR are to be reported
by stating that there are no changes. One
commenter felt that requiring a general
licensee to maintain its own FSAR (i.e.,
potentially separate and distinct from
the certificate holder) was unnecessary
and would cause confusion. One
commenter felt that the process for
revising the FSAR for a general licensee
was confusing.
The NRC agrees that providing a 24-
month FSAR update and adding the 6-
month cutoff for bringing the FSAR up
to date for changes made are consistent
with §50.71(e), are appropriate, and are
a reduction in unnecessary regulatory
burden. Lastly, the NRC believes that
providing a written confirmation when
no changes to the FSAR have been made
provides a clear and timely record of the
status of the FSAR to both the staff and
the public and agrees with this
comment. The NRC also agrees that
having a general licensee keep a
separate FSAR from that of a certificate
holder is redundant and believes that
requiring a separate FSAR is not
necessary for the staff to maintain its
regulatory oversight over general
licensees. Accordingly, proposed
paragraph (d) to §72.216 has been
withdrawn. In withdrawing this section,
the NRC wishes to clarify that the
certificate holder is not expected to
incorporate §72.48 changes made by
general licensees into its FSAR; rather
the certificate holder is responsible for
updating the FSAR for any changes it
has made under the provisions of
§72.48. Furthermore, the NRC expects
certificate holders to maintain the FSAR
current for any version of its cask
design, which is being used to store
spent fuel.
Two commenters suggested that the
proposed rule language in §§72.70, and
72.248 that the FSAR update include a
``description and analysis of changes in
procedures or in [SSC]'', was more
burdensome than the existing language
in §50.71(e) that the update is to
``contain all the changes necessary to
reflect information and analyses
submitted. * * *''
The NRC agrees that this language
could be read as requiring a separate
discussion of the effects of changes
beyond the SAR updates themselves,
which was not the intent of the
proposed rule. The language in §§72.70
and 72.248 has been revised to be as
consistent with §50.71(e) as possible
and, in particular, refers to ``include the
effects of'' changes, analyses and
evaluations, but not stating that the
update needs to describe each change.
In the current rule, a licensee must
submit to the NRC its FSAR 90 days
prior to the receipt of fuel or high level
waste and this action serves as a formal
notification to the regulator that fuel (or
high level waste) is planned to be
loaded. A number of comments viewed
this requirement as overly restrictive
because many changes related to cask
loading included in a FSAR will not be
identified or analyzed until
preoperational testing is performed and,
thus, the 90 day FSAR update
requirement could be interpreted as
another holdpoint before loading. The
NRC agrees that the requirement that a
FSAR be submitted at least 90 days
prior to fuel load was not intended to
serve as a holdpoint and in the final
rule, this has been changed to require a
specific licensee to submit a FSAR 90
days after receiving a license. To
maintain the notification aspect of the
current regulation, a new requirement
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was added to §72.80(g) to notify the
NRC of the licensee's readiness to begin
operation at least 90 days prior to the
first loading of spent fuel or high-level
radioactive waste. Specific licensees
will update their FSAR every two years.
Because the FSAR will be submitted
before construction and preoperational
testing of the ISFSI would be completed,
a requirement was retained in §72.70 to
provide a final analysis and evaluation
of the design and performance of SSCs
taking into account information since
the submittal of the application (i.e.,
information developed during final
design, construction, and preoperational
testing), in the next periodic update to
the FSAR. This information is not
required by the final §50.71(e);
however, it is necessary to require these
actions to complete the description of
the ISFSI, because of the single-step
licensing process in part 72.
New reporting requirements for
certificate holders will be added in
§§72.244 and 72.248, similar to existing
requirements imposed on licensees in
§§72.56 and 72.70, respectively.
4. §§72.3, 72.9, 72.24, 72.56, 72.86, and
72.212 (Miscellaneous Sections of Part
72)
No specific comments were received
on §§72.3, 72.9, 72.24 and 72.86, and
the final rule language is unchanged
from the proposed rule language for
these sections.
Two commenters believed that §72.56
was not clear on whether this regulation
applied to specific licensees, general
licensees, or both.
The NRC agrees and has revised this
section to indicate it applies to specific
licensees only.
One commenter suggested that §72.56
be revised to allow licensees to apply
for emergency or exigency processing of
license amendment requests, similar to
that allowed under certain conditions
for Part 50 licensees under §50.91(a)(5)
and (6).
The NRC disagrees. The NRC
currently has the authority under
§72.46(b)(2) to immediately issue an
amendment to a part 72 license upon a
finding that no genuine issue exists that
could adversely affect public health and
safety. Consequently, the NRC's
authority to immediately issue an
amendment to a part 72 license obviates
the need for a separate emergency or
exigency amendment process.
One commenter recommended that
any changes to the written evaluations
performed by a general licensee in
accordance with §72.212(b), in
determining whether a spent fuel
storage cask design can be used at a
particular part 50 reactor site, should be
accomplished using the requirements of
§72.48.
The NRC agrees and has revised
§72.212(b)(2)(ii) to require the general
licensee evaluate any changes to the
written evaluations required by §72.212
using the requirements of §72.48(c).
O.2 Petition for Rulemaking (PRM±72±3)
The NRC received a petition for
rulemaking submitted by Ms. Fawn
Shillinglaw in the form of two letters
addressed to Chairman Jackson dated
December 9 and December 29, 1995.
The Office of General Counsel
determined on March 5, 1996, that the
issues presented in these letters would
be treated as a petition for rulemaking.
The petition requested that the NRC
amend its regulations in 10 CFR part 72,
``Licensing Requirements for the
Independent Storage of Spent Fuel and
High-Level Radioactive Waste.'' The
petition was docketed as PRM±72±3 on
March 14, 1996. Ms. Shillinglaw
supplemented her petition with
additional information in a letter dated
April 15, 1996. The NRC published in
the Federal Register on May 14, 1996,
a notice of receipt of this petition and
stated the issues contained in the
petition (61 FR 24249).
Specifically, the petitioner requested
that the NRC amend those regulations
which govern independent storage of
spent nuclear fuel in dry storage casks
to require that: (1) The safety analysis
report (SAR) for a dry storage cask
design fully conforms with the
associated NRC safety evaluation report
(SER) and Certificate of Compliance
(CoC) before NRC certification (i.e.,
approval) of the dry storage cask design;
(2) the revision date and number of an
SAR be specified whenever that report
is referenced in documents; (3) the NRC
clarify the process for modification of an
SAR after a cask has been certified; and
(4) the NRC make available to the
public, the licensees' unloading
procedures. In her supplemental letter,
the petitioner recommended that to
eliminate confusion, the term ``CSAR''
(i.e., cask safety analysis report) be used
when referring to the SAR for any dry
storage cask design which has been
approved by the NRC and issued a CoC.
The Commission received ten
comment letters on PRM±72±3. The
commenters included five members of
the public, three public interest groups,
and the Nuclear Energy Institute (NEI).
Copies of the public comments on
PRM±72±3 are available for review in
the NRC Public Document Room, 2120
L Street, NW (Lower Level),
Washington, DC 20003±1527. No
comments were received objecting to
the petition. Eight of the commenters
were supportive of all, or some, of the
four issues raised in PRM±72±3. One
commenter (NEI), neither supported nor
opposed the petition and recommended
that any rulemaking action based on the
petition be delayed until the NRC
addressed issues in 10 CFR part 50
relating to the use of the ``FSAR'' as a
licensing basis document and the
application of §50.59 in 10 CFR part 50.
One commenter objected to NEI's
recommendation to delay rulemaking on
PRM±72±3.
The Commission has determined that
PRM±72±3 issues (1), (2), and (3) should
be granted, in part; and issue (4) should
be denied. This notice constitutes the
Commission's final action on this
petition. The basis for the Commission's
actions on each issue and responses to
public comments received on the
petition are described below.
Issue (1): Part 72 should be amended
to require that the safety analysis report
(SAR) for a spent fuel dry storage cask
design fully conforms with the
associated NRC safety evaluation report
(SER) and certificate of compliance
(CoC) before NRC certification (i.e.,
approval) of the cask design.
Five comment letters were received
supporting Issue (1) of PRM±72±3.
Resolution of Issue (1): In this final
rule the Commission has granted, in
part, the petitioner's request on this
issue. This rule adds new §72.248 to
part 72 and this section addresses this
issue by requiring a certificate holder to
submit a final safety analysis report
(FSAR) after issuance of the CoC. This
rule also describes the process for
periodic updates of the FSAR. Section
72.248, paragraphs (a)(1) and (a)(2) state,
in part:
Each certificate holder shall submit an
original FSAR to the Commission ** *
within 90 days after the spent fuel storage
cask design has been approved pursuant to
§72.238. This original FSAR shall be based
on the safety analysis report submitted with
the application and reflect any changes and
applicant commitments developed during the
cask design review process. The original
FSAR shall be updated to reflect any changes
to requirements contained in the issued
Certificate of Compliance (CoC). ** *
The Commission agrees with the
petitioner that the FSAR should be fully
conformed (i.e., consistent) with the
operating limits contained in the CoC,
because the FSAR contains the design
information the staff used to make its
safety finding and to approve the dry
storage cask design for use. The
Commission disagrees with the
petitioner's request that the FSAR be
conformed to the NRC SER for the dry
storage cask design, and that the FSAR
be submitted to the NRC before approval
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of the cask design (i.e., issuance of the
CoC). The NRC SER contains staff
conclusions on the adequacy of the cask
design, not applicant commitments to
the NRC on the cask design. Therefore,
the Commission believes it is not
necessary to conform the FSAR to the
issued NRC SER before the CoC can be
issued. The NRC SER is available in the
NRC Public Document Room for public
review.
The Commission disagrees with the
petitioner's request that issuance of the
CoC (i.e., placement of the CoC in the
list at §72.214 which enables a general
licensee to use the cask design) be
delayed until after the certificate holder
has submitted an FSAR to the NRC (i.e.,
updated the topical safety analysis
report, submitted with its application
for approval of a dry storage cask
design, to ensure that the SAR is
consistent (fully conforms) with the
approved CoC). This final rule codifies
as a regulation the NRC's current
approach which, administratively,
requires a certificate holder to update its
SAR after issuance of the CoC to ensure
it is consistent with the issued CoC. For
administrative purposes, the
Commission prefers that the original
FSAR be submitted to the NRC, within
90 days after the CoC is issued, so that
the certificate holder can include
[conform] in the FSAR any conditions
from the issued CoC. The FSAR does
not need to be conformed to the CoC,
before the CoC is issued, because this
action does not provide any new
information the NRC would need to
make a determination that the cask
design meets the requirements of part
72, subpart L, and is acceptable for use.
The Commission also disagrees with
the petitioner's supplemental
information to use the term ``cask safety
analysis report (CSAR)'' when referring
to the SAR submitted after the NRC
approves a cask design. Instead, the
Commission is using the term ``final
safety analysis report (FSAR)'' to
identify the SAR submitted after the
NRC approves a cask design. The use of
the term ``FSAR'' is the accepted
practice by industry and will not cause
confusion. Further, this approach will
ensure consistency between parts 50
and 72, because the term ``FSAR'' is
used by §§50.59, 50.71(e), 72.48, and
72.70 in this final rule.
Issue (2): Part 72 should be amended
to require that the revision date and
number of an SAR be specified
whenever that report is referenced in
documents.
Five comment letters were received
supporting Issue (2) of PRM±72±3.
Resolution of Issue (2): In this final
rule the Commission has granted, in
part, the petitioner's request on this
issue. This rule adds new §72.248 to
part 72 which requires that revision
numbers, change bars, and a list of
current pages be included in any
revisions to the FSAR. Section 72.248,
subparagraphs (c)(2) and (c)(3) state:
The update [of the FSAR] shall include a
list that identifies the current pages of the
FSAR following page replacement. Each
replacement page shall include both a change
indicator for the area changed, e.g., a bold
line vertically drawn in the margin adjacent
to the portion actually changed, and a page
change identification (date of change or
change number or both).
These features will clearly identify
what has been changed, as well as the
date of the change, in any revision to a
FSAR. While §72.248 will provide a
process for requiring revisions to the
FSAR be clearly indicated, the
Commission has denied the portion of
the petitioner's request to amend part 72
to require a FSAR revision number and
date be specified when the FSAR is
referenced in other documents (e.g., an
application for a part 72 license or CoC).
Instead, the NRC will revise guidance
documents for part 72 activities (e.g.,
regulatory guides and standard review
plans) to require specification of the
FSAR revision date and number
whenever a FSAR is referenced in
another document. The Commission
believes addressing this portion of the
petitioner's request in guidance
documents rather than in a regulation is
more appropriate and meets the intent
of the request.
Issue (3): The NRC must clarify the
process for modification of a safety
analysis report after a cask [design] has
been certified (i.e., approved by the
NRC).
Five comment letters were received
supporting Issue (3) of PRM±72±3
including a comment from the
petitioner clarifying that she believed
that ``any changes to the SAR (FSAR)
should be done by the amendment
process of rulemaking.'' Four
commenters also recommended that any
changes made to the SAR (including a
generic SAR), the cask design, or the
CoC should require rulemaking and
public comment or a public hearing.
One commenter also suggested that the
regulations be amended to include more
detail on who can make changes to dry
storage cask designs and whether
vendors (i.e., certificate holders) can
make these changes.
Resolution of Issue (3): The
Commission is revising §72.48 to allow
a certificate holder to make certain types
of changes to a cask design, or
procedures, or to conduct tests and
experiments, not described in the FSAR
(as updated) without requiring prior
NRC approval if the criteria in §72.48(c)
are met. If these criteria are not met, a
certificate holder must obtain a CoC
amendment pursuant to §72.244.
Following such changes (either resulting
from the §72.48 process or the CoC
amendment process), the certificate
holder must update the FSAR as
required by §72.248. Section 72.248,
paragraphs (b), (b)(2), and (b)(3) state, in
part:
The (FSAR) update shall include the
effects of: All safety analyses and evaluations
performed by the certificate holder either in
support of approved CoC amendments, or in
support of conclusions that the changes did
not require a CoC amendment in accordance
with §72.48. All analysis of new safety issues
performed by or on behalf of the certificate
holder at Commission request. The
information shall be appropriately located
with the updated FSAR.
The Commission is seeking to reduce
any unnecessary regulatory burden
placed on its licensees and certificate
holders without compromising safety.
The dry storage cask design review
process and the analysis acceptance
criteria are defined in the NRC's
standard review plans. This final rule
allows licensees and certificate holders
to make changes to the cask design,
without obtaining prior NRC approval,
for changes which do not significantly
impact the ability of the cask to perform
its intended functions. The impact of
these changes are then incorporated into
an updated FSAR, which is submitted to
the NRC. Requiring that all changes to
a cask design or changes to a FSAR be
reviewed and approved by the NRC
through the rulemaking amendment
process, including either a public
comment period or a public hearing,
defeats these efforts with no discernable
increase in safety. Further, while
rulemaking is currently utilized to
amend a CoC, the Commission is
presently re-examining the
appropriateness of this procedure.
Therefore, the Commission has granted
petitioner's request to clarify the process
for modification of an FSAR after the
NRC has approved the cask design and
issued the CoC, but has rejected the
request to require all changes to a cask
design, or the FSAR, be made via a
rulemaking amendment process.
Issue (4): The NRC should make cask
unloading procedures publicly
available.
Five comment letters were received
supporting Issue (4) of PRM±72±3. One
commenter also requested that the NRC
review, approve, and have tested
unloading procedures prior to their
being implemented. One commenter
suggested suspending all cask loading
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activities until the NRC reviews
procedures [for loading and unloading]
and appropriate tests are completed.
Resolution of Issue (4): The NRC does
not approve or test a licensee's loading
or unloading procedures, rather the
licensee is responsible for development,
verification, and validation of the
loading and unloading procedures. The
NRC inspects the licensee's procedures
(i.e., reviews the procedures and
observes the licensee implementing
them) to determine whether the
procedures will provide reasonable
assurance that public health and safety
will be adequately protected.
The Commission does not agree that
cask unloading procedures should be
required to be public documents. First,
in order to make these procedures
publicly available, either the NRC must
possess the procedures, or the licensee
must place the procedures in the public
domain. The Commission's position is
that only those documents necessary to
demonstrate that a dry storage cask is
designed to meet the requirements of
part 72, subpart L, need to be submitted
to the NRC on the docket (i.e., to allow
the NRC to determine that the cask
design is acceptable for use). Cask
loading and unloading procedures are
implementing documents required by
the CoC which are developed and
implemented by the licensee.
Although the NRC does not possess
the procedures, they are subject to
inspection by NRC staff. However, even
during inspection activities, NRC
generally does not take possession of the
procedures. Therefore, the unloading
procedures remain the property of the
licensees and are not available to the
public. The NRC's inspection program
for part 72 licensees requires the
inspection of loading and unloading
activities, including a review of
applicable procedures, before a licensee
begins cask loading. NRC inspection
personnel perform these activities at the
licensee's site and observe the licensee's
preoperational testing and dry run
activities to assess the adequacy of these
procedures and the readiness of the
licensee to begin loading spent fuel. The
results of these inspections are
documented in reports which are placed
in the NRC Public Document Room and
are available for public review.
Furthermore, requiring part 72
licensees to submit their implementing
procedures to the NRC (i.e., operating
procedures such as loading and
unloading procedures, maintenance
procedures, surveillance procedures,
radiation protection procedures,
security procedures, emergency
procedures, and administrative
procedures), as well as any revisions to
these procedures, would impose a huge
paperwork burden on both the licensee
and on NRC staff without a
corresponding safety benefit. Therefore,
Issue (4) is denied.
Additional Public Comments on the
Petition
In addition to the specific comments
that were received on the petition that
are discussed above, a number of
comments were received on related and
unrelated subjects.
Comment: Five comments were
received on the VSC±24 cask design
being used at the Palisades and Point
Beach plants and incidents related to
the VSC±24 cask design.
Response: The Commission considers
these comments beyond the scope of
this petition and this rulemaking.
Comment: Two comments were
received suggesting that when a change
to an approved dry storage cask design
is requested, that the existing CoC be
suspended until the changes are
approved by the NRC.
Response: The Commission considers
these comments would impose an
unreasonable burden on part 72
licensees. Suspending a CoC solely on
the basis of receiving a change and not
on the basis of a compelling safety need,
would imply that any casks
manufactured under the CoC, which are
in use by part 72 licensees, should be
taken out of service (i.e., unloaded)
upon receipt of any request to revise the
cask design. Requiring that a cask be
unloaded in these circumstances would
impose an unreviewed backfit on the
part 72 licensees using that cask design
and would also result in unnecessary
occupational exposure to licensee
workers.
Comment: One comment was received
recommending that any rulemaking
action based on PRM±72±3 be delayed
until the NRC addressed issues in 10
CFR part 50 relating to the use of the
``FSAR'' as a licensing basis document
and the application of §50.59 in 10 CFR
part 50. Another commenter disagreed
with this recommendation to delay
rulemaking on PRM±72±3.
Response: The Commission believes
that issuance of this final rule resolves
this comment.
Comment: One commenter requested
that the NRC prohibit general licensees
from using §72.48 and only permit cask
design changes via rulemaking. One
commenter recommended that any
identification of an unreviewed safety
question submitted to the NRC should
require that NRC conduct a hearing on
the issue. One commenter suggested
that the NRC approve each §72.48
safety evaluation and place each
evaluation in the public document
room. One commenter suggested that
the NRC ``vacate the generic ruling
procedure'' subpart L and require that
public hearings be held prior to NRC
cask certification. One commenter
suggested a moratorium on additional
dry cask storage cask designs.
Response: Petitioner's concerns
related to cask certification issues; in
particular, the process for modifying a
SAR for a dry cask storage design before
and after issuance of the CoC. These
comments raise broad policy issues that
go well beyond the scope of this petition
and rulemaking.
O.3 Part 71 (Transportation) Comments
Several commenters stated that a
change control process similar to §72.48
should be established in part 71 for
transportation. These commenters noted
that for dual-purpose casks, used for
both transportation and storage, the lack
of a process in part 71 would limit the
usefulness of the authority provided
under §72.48. Although the
Commission agrees that this comment
has merit, adding this authority to part
71 is beyond the scope of the proposed
rule. In response to these comments, the
Commission will consider adding
``§71.48-type'' change authority as part
of a currently planned rulemaking for
part 71 intended to update requirements
for compatibility with the most recent
International Atomic Energy Agency
transportation standards.
P. Other Topics Discussed in the Notice
and Comments Not Related to Preceding
Topic Areas
The Federal Register notice
containing the proposed rule also
solicited comments on particular topics
that were discussed in the preceding
sections. In addition, comments were
received on a number of aspects not
directly related to the rule language
itself, such as guidance, enforcement
policy, the regulatory (and backfit)
analysis, or on other issues.
Guidance
Many comments were received on the
subject of guidance. Many suggested
that NEI and NRC work together to
develop guidance, and that the guidance
be endorsed before the revised rule
becomes effective. Commenters also
requested examples of such matters as
interdependent changes, minimal
increases, and screening of changes (as
discussed in Sections B and G).
The NRC agrees that guidance is
important, and notes that NEI has stated
its willingness to revise existing
guidance to conform with the final rule
such that NRC could endorse it. The
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NRC will work with interested
stakeholders to agree upon guidance
that includes consideration of these
issues. Further, NRC is delaying the
required implementation of the rule for
several months to allow time for
guidance to be revised.
Fuel Burnup Limits
One commenter stated that NRC
should clarify the acceptance limits of
§51.55 concerning burnup assumptions
for the transportation of spent fuel for
BWRs, as well as clarifying if this is
subject to §50.59 evaluations.
The Commission notes that a
proposed rule (§51.52, not §51.55 as
cited by the commenter) was recently
published on February 26, 1999 (64 FR
9884), concerning environmental
implications of higher burnup fuel for
transportation of spent fuel.
Transportation of fuel is not covered by
§50.59 (as noted elsewhere in this
notice, the Commission is considering
revisions to part 71 that would add a
change control process similar to §50.59
that could be used for changes to
transportation requirements under part
71). If the commenter was asking
whether higher burnup fuel can be used
without NRC approval, it is unlikely
that such a change would satisfy the
criteria of §50.59, either because TS
changes would be involved, other
requirements (e.g., §50.46) would not be
met, or the burnup being considered
would be outside the range of what was
approved in the topical reports for the
fuel.
Alternative Criteria
Two commenters proposed the use of
alternate criteria for reactors that are
being decommissioned. One commenter
suggested that a ``margin'' criterion is
not necessary, but that a criterion on
environmental impact might be
appropriate.
The Commission notes that the new
criteria in the final rule that replace the
``margin'' criterion are appropriate for a
reactor being decommissioned. Further,
§50.82(a)(6) specifies that licensees
shall not perform any decommissioning
activities that result in significant
environmental impact not previously
reviewed. Section 50.82(a)(4) requires
that the post-shutdown
decommissioning activities report
include a discussion that provides the
reasons for concluding that the
environmental impacts associated with
site-specific decommissioning activities
will be bounded by appropriate,
previously issued environmental impact
statements. For these reasons, the
Commission concludes that a criterion
on environmental impact is not needed.
The second commenter stated that the
scope of §50.59 should be limited to
systems related to spent fuel pool
cooling or radiological waste.
The Commission notes that the staff
involved in requirements for
decommissioning are developing
guidance on the scope of information
required to be in an updated FSAR for
a reactor undergoing decommissioning.
This effort is examining what
information should be retained in an
FSAR for these facilities. The
Commission believes that defining the
scope of information required to be in
the FSAR for a reactor undergoing
decommissioning would be the best way
to address the apparent concern raised
in this comment, rather than by
modifying §50.59 as recommended by
the commenter.
Regulatory Analysis
Some comments were received on the
regulatory analysis, primarily that NRC
underestimated the impacts on NRC and
licensees of the number of license
amendments that would result, or the
burden on part 72 licensees. These
comments would appear to reflect a
view that the proposed rule would
require more amendments than are
currently required, perhaps because of
differences between the proposed rule
language and existing practice of some
licensees using NEI 96±07, or depending
upon which formulation of ``margin of
safety'' was ultimately adopted. The
Commission has prepared a final
regulatory analysis that reflects the final
rule language and consideration of the
public comments. The Commission does
not agree that the final rule language
will result in more amendments than
presently arise under the existing rule.
Need for Further Notice and Comment
Two commenters stated that the
Commission should ensure that the final
rule is within the bounds of the
proposed rule notice, or should provide
opportunity for public comment on
substantive changes. The Commission
has examined the final rule for
consistency with the proposed rule and
concludes that the final rule is within
the bounds of the proposed rule, taking
due consideration of the public
comments that sought clarification and
revisions in some respects, as well as
greater consistency between the Part 50
and Part 72 requirements.
Different Process for non-TS Issues
Several commenters believe that the
license amendment process is not well
suited to the type of changes that
require review under §50.59(c)(2), but
that do not involve changes to the TS or
the license directly. They believe that
the Commission should establish a
different review process for such
changes, such as letter approval.
The Commission notes that at one
time (until 1974), §50.59 did contain
two approval processes, one for license
amendments, and the other for
``authorizations.'' The rule was revised
in 1974 to delete the ``authorization''
process and to handle all the required
approvals as license amendments. The
Commission notes that the present
rulemaking provides some relaxation in
the evaluation criteria. Therefore, the
NRC has responded to concerns about
having to process a license amendment
for ``minimal'' changes. The current
process provides opportunity for public
participation in the process under the
provisions of §50.90 for changes that
exceed the criteria, and for public
knowledge, through the summary
reports, of those matters that did not
require prior approval. Therefore, the
Commission does not plan to establish
a different process.
Other Definitions
Some commenters felt that NRC
should provide better definitions of
certain terms that appear in §50.59 (and
elsewhere), specifically, for ``design
bases'' and for ``important to safety.''
The Commission notes that §50.2
does define design bases, but also notes
that efforts are underway within the
agency to enhance understanding of
what constitutes design basis
information, through possible
development of criteria and examples.
Concerning ``important to safety,'' the
Commission does not believe that a
definition is critical to implementation
of the rule, since the set of SSCs viewed
as important to safety was arrived at
during the license review and are
described in the FSAR. Thus, lack of an
established definition is not an
impediment to implementation of the
rule (the Commission notes that for part
72, a definition is provided for SSC
important to safety).
Applicability to Part 76
In its development of the proposed
rule, as discussed in SECY±98±171, the
staff recommended exclusion of part 76
(``Certification of Gaseous Diffusion
Plants'') from those regulations for
which rule changes were being
proposed. The basis for this
recommendation was a lack of design
detail currently available in the safety
analysis reports for these plants. One
commenter argued that the flexibility
provided by the revised evaluation
criteria should also be included in
§76.68 (this section contains
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requirements very similar to existing
§§50.59 and 72.48). This commenter
stated that the process by which
changes are evaluated should not vary
based on the detail of the description
being changed.
The Commission notes that the
gaseous diffusion plants (GDP) have
significantly less design basis
information than is currently available
for reactor facilities. The lack of design
detail and lack of understanding of the
design basis has been documented in
the Compliance Plans for the GDPs, in
NRC inspection reports, and is evident
in the GDP SARs. The Commission
concludes that successful
implementation of a change control
process is dependent upon the level of
knowledge about the design basis of the
plant equipment or operation being
changed. At the present time, the
Commission does not believe that
additional flexibility is appropriate for
part 76 facilities.
Q. Enforcement Policy
Some commenters raised issues about
how enforcement decisions would be
made during the transition period, and
following implementation, particularly
with respect to evaluations performed in
the past.
The Commission recognizes that it
will take time to revise existing industry
guidance and to revise procedures, and
conduct training on the new rule
provisions before the rule can be fully
implemented. There will still be the
possibility of finding previous plant
changes performed prior to the
implementation of the new rule that
would be potential violations of the
previous rule. The Commission has
concluded that enforcement of potential
violations of §§50.59 and 72.48 for past
evaluations will be handled as
described below, and also in accordance
with the NRC Enforcement Policy,
NUREG±1600, Revision 1.
Following publication of the revised
rule, for situations that violate the ``old''
requirements, but that would not be
violations had the evaluation been
performed under the revised rule, the
NRC will exercise enforcement
discretion pursuant to VII.B.6 of the
Enforcement Policy and not issue
citations against the ``old'' rule. The
staff will document in inspection
reports that the issue was identified, but
that no enforcement action is being
taken because the revised rule
requirements are met. However, for
those situations identified prior to the
effective date of the revised rule that
involve a violation of the existing rule
requirements but that would not be
violations under the revised rule,
licensees still need to take the required
corrective action within a reasonable
time frame commensurate with safety
significance to avoid the potential for a
willful violation of NRC requirements.
The NRC plans to maintain an
enforcement panel made up of NRR
(and NMSS as applicable), OE, and OGC
representatives for some months after
publication to maintain consistency.
Additional enforcement policy changes
that may be applicable to violations of
§§50.59 or 72.48 are under
consideration. The Commission intends
to revise NUREG±1600, Rev. 1, ``General
Statement of Policy and Procedures for
NRC Enforcement Actions,'' consistent
with this enforcement approach prior to
the effective date of the rule.
R. Implementation
The Commission recognizes the role
that regulatory guidance will play in
effective implementation of the
revisions to the rule. Existing guidance
(e.g., NEI 96±07 and NRC inspection
guidance) needs to be revised to
conform with the rule changes. To allow
time for the guidance to be revised, and
for licensees to implement the revised
rule provisions using the revised
guidance, the Commission has
established that the rule changes to part
50 will become effective 90 days after
promulgation of the final regulatory
guidance.
For part 72 facilities, current
schedules for guidance would result in
availability at a time later than that
anticipated for the guidance for part 50.
Accordingly, the effective date for these
sections is longer, set at 18 months from
publication of the rule in the Federal
Register. For those sections in part 72
for which no guidance is needed, as for
instance, §§72.244 and 72.246, the
effective date is 120 days from
publication.
III. Section by Section Analysis
10 CFR Part 50
10 CFR 50.59
As discussed in more detail above,
§50.59 is being restructured and revised
to have the following components:
Paragraph (a): This is a new
paragraph that contains definitions of
terms used in the rule. The terms
establish requirements for when
evaluations are to be conducted to
determine if the proposed changes,
tests, or experiments meet the criteria to
require prior NRC approval.
Accordingly, definitions are given for
``change,'' ``facility as described in the
final safety analysis report (as updated)
* **,'' ``procedures as described
* **,'' ``tests and experiments not
described * **'' etc. The specific
definitions were discussed in the
preceding sections.
Paragraph (b): Relocation into one
paragraph of existing applicability
provisions. Section 50.59 applies to
facilities licensed under part 50,
including power reactors and non-
power reactors, whether operating or
being decommissioned.
Paragraph (c)(1): Relocation and
clarification of existing provisions
establishing which changes, tests, or
experiments require evaluation and
process for receiving approval when
necessary. The provisions now use the
terms defined in paragraph (a), and refer
to the ``final safety analysis report (as
updated),'' rather than to ``safety
analysis report.'' The terminology of
``unreviewed safety question'' has been
replaced by referring to the need to
obtain a license amendment.
Paragraph (c)(2): Reformatting of the
(existing) evaluation requirements into
seven distinct statements of the criteria,
addition of an eighth criterion, and
revision of the existing criteria for when
prior NRC approval of a change, test, or
experiment is required. Specifically,
language of ``more than a minimal
increase in frequency (or likelihood),''
and of ``more than a minimal increase
in consequences'' was inserted in the
criteria concerning accidents and
malfunctions, and rule requirements
were revised from ``may be created'' to
``would create'' concerning creation of
accidents of a different type and
malfunctions of structures, systems, and
components important to safety with a
different result (instead of existing
language of malfunction of equipment of
a different type). In addition, the
existing criterion on ``margin of safety''
was replaced by a criterion focusing
upon design basis limits for fission
product barriers being exceeded or
altered, and a new criterion was added
to control evaluation methods. These
revisions clarify the criteria for when
prior approval is needed and allow
some flexibility for licensees to make
changes that would not affect the NRC
basis for licensing of the facility.
Paragraph (c)(3): This is a new
paragraph containing the requirement
that evaluations and analyses performed
since the last FSAR update was
submitted need to be considered in
performing evaluations of changes to the
facility or procedures, or for conduct of
tests and experiments. This paragraph is
consistent with the terminology of
``final safety analysis report (as
updated).''
Paragraph (c)(4): This is a new
paragraph that states that §50.59
requirements do not apply to changes to
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the facility or procedures when other
regulations establish more specific
criteria for such changes. Thus, this
paragraph clarifies that duplicative
reviews in accordance with §50.59 are
not necessary for information that is
described in the FSAR, but for which
other regulations provide standards for
change control.
Paragraph (d)(1): Renumbered
paragraph with (existing) recordkeeping
requirements. The text was simplified
concerning which records are needed,
and conforming changes were made for
the change in terminology from ``safety
evaluation'' to ``evaluation.''
Paragraph (d)(2): Renumbered
paragraph with (existing) reporting
requirements. The text was simplified to
state that summary reports must be
submitted at least once every 24
months, instead of the existing
statement that refers to submitting the
summary report along with the FSAR
update submittal or annually. This
revision will allow all facilities to
submit the report on a 24 month
frequency.
Paragraph (d)(3): Renumbered
paragraph on retention of records. The
text was revised to cover retention of
records required by §50.59 until the
term of any renewed license has
expired.
10 CFR 50.66
This section specifies requirements
for thermal annealing of a reactor
pressure vessel. The changes to §50.66
are to conform existing language
referring to unreviewed safety
questions, and to updated final safety
analysis report, to the language in
revised §50.59.
10 CFR 50.71(e)
This section discusses requirements
for periodic updating of the final safety
analysis report, to reflect the effects of
changes made either under §50.59, or
through license amendments, or effects
of new analyses. The changes to this
section are to conform language with
respect to unreviewed safety question,
safety evaluation, and reference to the
final safety analysis report (as updated),
with the language in revised §50.59, as
well as other minor wording changes as
noted above (e.g., ``approved'' license
amendments).
10 CFR 50.90
A portion of existing §50.59(c) is
being relocated into this section. This
change places the requirements for
changes to technical specifications
themselves (not a result of a change, test
or experiment as defined in §50.59),
into the rule section on amendments to
licenses rather than retaining the
requirement in the section on changes to
the facility.
10 CFR Part 72
Most of the revisions in part 72 mirror
those made to §50.59. As for part 50,
other changes are needed with respect
to updating of safety analysis reports,
and in other sections for consistent
terminology.
10 CFR 72.3
The definition of ``independent spent
fuel storage installation'' is being
revised to remove the tests for
evaluation of the acceptability of
sharing common utilities and services
between the ISFSI and other facilities.
(Section 72.24 is being revised to
include this evaluation.)
10 CFR 72.9
Paragraph (b) is being revised as a
conforming change to include in the list
of information collection requirements
the new requirements in §§72.244 and
72.248 for amendments and for updates
to the safety analysis reports by CoC
holders.
10 CFR 72.24
This section is being revised to
reference shared common utilities and
services in the applicant's assessment of
potential interactions between the ISFSI
and another facility (previously covered
by §72.3).
10 CFR 72.48
This section is being totally
reformatted and revised, as discussed
above for §50.59. Specifically, it
contains the following:
Paragraph (a): This paragraph now
specifies definitions for terms such as
``change'' and ``facility as described in
the Final Safety Analysis Report (as
updated).'' Additionally, the term
``Final Safety Analysis Report (FSAR)
(as updated)'' has been defined to
provide greater clarity and consistency
with §50.59 and other sections of part
72.
Paragraph (b): This paragraph
specifies that this section is applicable
to general and specific licensees for an
ISFSI or MRS, and to spent fuel storage
cask certificate holders.
Paragraph (c): Paragraph (c)(1)
establishes the conditions a licensee or
certificate holder must meet in order to
(1) make changes to the facility or spent
fuel storage cask design as described in
the FSAR, or (2) make changes to the
procedures as described in the FSAR, or
(3) conduct tests or experiments not
described in the FSAR, without prior
NRC approval. Those conditions are
that: (1) A change to the technical
specifications is not required; (2) a
change in the terms, conditions or
specifications incorporated in the CoC is
not required; and (3) the change, test, or
experiment does not meet any of the
criteria in paragraph (c)(2).
Paragraph (c)(2) lists the specific
criteria which, if met, permit a licensee
or certificate holder to make the
changes, or conduct the tests or
experiments, described in paragraph
(c)(1) without NRC approval. These new
criteria revise existing criteria and
conform with the criteria adopted in
§50.59(c)(2). Two existing criteria
involving a significant increase in
occupational exposure or a significant
environmental impact have been
deleted. Paragraph (c)(3) states that
changes made but not yet reflected in
the FSAR update also need to be
considered in making the determination
under paragraph (c)(2). Paragraph (c)(4)
states that §72.48 does not apply to
changes to the facility or procedures
when the regulations establish other
change control processes for such
changes.
Paragraph (d): This paragraph
contains the recordkeeping
requirements and reporting
requirements. In the final rule,
subsection numbers were included for
clarity. For records, the rule is revised
to refer to the records of determinations
of the need for license or certificate of
compliance (CoC) amendments, rather
than to records involving unreviewed
safety question determinations. The
time frame for submitting summary
reports in (renumbered) paragraph (d)(2)
was revised from 12 months to 24
months. The filing requirements for the
summary reports are modified to be
consistent with §72.4
(Communications).
Paragraphs (d)(3), (d)(4) and (d)(5)
contain record retention requirements.
The retention requirements for changes
to procedures and conduct of tests and
experiments were revised to be 5 years
(instead of until termination). These
time frames are more consistent with
those in §50.59, and also reflect that
while facility changes need to be
maintained until termination, other
records are of less importance after a
period of time such as 5 years.
Paragraph (d)(3)(i) and (d)(3)(ii) are
renumbered and clarified with respect
to when records no longer need to be
maintained.
New paragraph (d)(6) requires
licensees who make changes under
§72.48 to provide copies of the records
of such changes to the certificate holder
for the cask, and for the certificate
holders who make changes to provide
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53610 Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
records to the general and specific
licensees using that cask, within 60 days
of implementing the changes.
10 CFR 72.56
Existing §72.48(c)(2) is being
relocated into this section. This is a
parallel change to that for §§50.59 and
50.90. The Commission is placing the
requirements for changes to license
conditions in the rule section on
amendments to licenses instead of in
the section on changes to the facility.
10 CFR 72.70
This section contains requirements for
updating of safety analysis reports by
licensees. Section 72.70 was reformatted
and revised to conform more closely
with the update requirements in
§50.71(e), as well as those in (new)
§72.248. The update frequency is being
revised from 12 months to 24 months.
Paragraphs (a) and (b) are being revised
to use the terms ``Final Safety Analysis
Report,'' ``FSAR,'' and ``as updated.''
Paragraph (a) is also being revised to
indicate the original FSAR for a specific
licensee will be submitted within 90
days of issuance of the license. Final
analyses associated with completion of
construction or preoperational testing
will be provided in the next periodic
update of the FSAR. The requirement
for a licensee to submit a FSAR 90 days
before planned receipt of spent fuel has
been removed, in lieu of a notification
under §72.80(g) by the licensee 90 days
before ISFSI operation commences. The
section is also being revised to add the
requirement that changes to procedures
be reflected in the periodic updates of
the FSAR. New paragraph (c) is being
added to provide requirements on
submitting revisions to the FSAR for
specific licensees, including provisions
for replacement pages, a cut off date for
changes, time frame to file, and
provisions for updating if no changes
were made.
10 CFR 72.80
New paragraph (g) is being added to
this section to require a specific licensee
to notify the NRC at least 90 days in
advance of its readiness to commence
ISFSI (or MRS) operations This
requirement replaces a requirement in
present §72.70(a) that an FSAR be
submitted to the Commission at least 90
days prior to the planned receipt of
spent fuel or high-level waste. This
requirement thus ensures that the NRC
is informed in advance of licensee plans
to use the facility so that appropriate
oversight activities can be conducted.
10 CFR 72.86
Paragraph (b) currently includes those
sections under which criminal sanctions
are not issued. This paragraph is being
revised to add §§72.244 and 72.246 as
a conforming change to reflect that
certificate holders who fail to comply
with these new sections would not be
subject to the criminal penalty
provisions of section 223 of the Atomic
Energy Act (AEA). New §72.248 has not
been included in paragraph (b) to reflect
that certificate holders who fail to
comply with this new section would be
subject to the criminal penalty
provisions of section 223 of the AEA.
10 CFR 72.212(b)(2)
Paragraph (b)(2)(i) retains the current
rule language but has been renumbered
and reordered for clarity as a result of
the addition of paragraph (b)(2)(ii).
Paragraph (b)(2)(ii) was added to require
that the general licensee evaluate any
changes to the written evaluations
required by §72.212 using the
requirements of §72.48(c).
10 CFR 72.212(b)(4)
The change to this section is to
conform the reference to §50.59
provisions, specifically to change from
the terminology of unreviewed safety
question to referring to the need for a
license amendment for the facility (that
is, the reactor facility at whose site the
independent spent fuel storage
installation is located).
10 CFR 72.216
In the proposed rule, a new paragraph
(d) would have been added to present
requirements for a general licensee to
submit annual updates to a final safety
analysis report (FSAR) for the cask or
casks approved for spent fuel storage
that are used by the general licensee. In
the final rule, this section was
withdrawn because the Commission
concluded that it was not necessary for
general licensees to submit updates to
the safety analysis report for the
approved cask design that they are using
for storage.
10 CFR 72.244
This new section presents
requirements for how a certificate
holder is to submit an application to
amend the certificate of compliance
(CoC). This section is similar to the
requirements in §72.56 for licensees to
apply for an amendment to their license.
10 CFR 72.246
This new section presents
requirements for approval of an
amendment to a CoC. This section is
similar to the requirements in §72.58
for approval of an amendment to a
license.
10 CFR 72.248
This new section presents
requirements for submittal of periodic
updates to an FSAR associated with the
design of a spent fuel storage cask
which has been issued a CoC. This new
section also states that the changes to
procedures and SSC associated with the
spent fuel storage cask and which are
made pursuant to §72.48 would be
included in the update. This section is
similar to the requirements in §72.70
for submission of updates to the FSAR
associated with a part 72 license and to
the requirements in §50.71(e) for power
reactor FSAR updates.
IV. Finding of No Significant
Environmental Impact
The Commission has determined
under the National Environmental
Policy Act of 1969, as amended, and the
Commission's regulations in subpart A
of 10 CFR part 51, that this rule, as
adopted, will not have a significant
impact on the environment. The rule
changes are of two types: those that
relate to the processes for evaluating
and approving changes to licensed
facilities and those that involve the
degree of potential change in safety for
which changes can proceed without
NRC review. The process changes will
make it more likely that planned
changes are properly reviewed and
approved by NRC when necessary. With
respect to the criteria changes, only
minimal increases in frequencies of
postulated design basis accidents will
be allowed without prior NRC review.
All changes to the Technical
Specifications, which are the operating
limits and other parameters of most
immediate concern for public health
and safety, will continue to require prior
NRC review and approval. Changes to
the facility that would involve an
accident of a different type from any
already analyzed require prior approval.
Further, changes that result in more
than minimal increases in radiological
consequences will continue to require
prior NRC approval, including NRC
consideration as to whether there is a
potential impact on the environment.
Therefore, the Commission concludes
that there will be no significant impact
on the environment from this rule. This
discussion constitutes the
environmental assessment and finding
of no significant impact for this
rulemaking.
V. Paperwork Reduction Act Statement
This rule amends information
collection requirements that are subject
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4In some cases, these changes coincide with other
changes intended to clarify and codify existing
practice, and to make the rule easier to understand
(e.g., separating the ``frequency of occurrence'' of an
accident from the ``consequences'' of an accident as
a criterion for NRC review and approval.
5``Permissive'' relaxations are relaxations which
licensees may voluntarily choose (but are not
compelled) to comply.
to the Paperwork Reduction Act of 1995
(44 U.S.C. 3501 et seq.). The proposed
rule was submitted to the Office of
Management and Budget for review and
approval of the information collection
requirements. Existing requirements
were approved by the Office of
Management and Budget approval
numbers 3150±0011 and 3150±0132.
The rule changes affect information
collection requirements through the
existing reporting requirements in
§50.59 for a summary report of changes,
tests and experiments, performed under
the authority of §50.59 as well as
recordkeeping requirements. Similar
requirements exist in §72.48 for
licensees under part 72. In addition,
revisions are being made to the
requirements in §72.70 and (new)
72.248 for submittal of updates to the
safety analysis reports. Further, the final
rule establishes recordkeeping and
reporting requirements for CoC holders
who make changes to an approved
storage cask design in accordance with
§72.48.
The public reporting burden for this
information collection request was
estimated in the proposed rule to
average 3100 hours per response,
including the time for reviewing
instructions, searching existing data
sources, gathering and maintaining the
data needed, and completing and
reviewing the information collection.
The Commission had estimated that
there would be only a slight increase in
burden associated with these proposed
changes over the existing burden. For
the final rule, certain of the provisions
that might have resulted in an increase
in burden have been removed; therefore,
the Commission now concludes that the
final rule would result in an overall
reduction in reporting and
recordkeeping burden, other than for the
estimated effort required for a one-time
revision to procedures and training.
Therefore, the present estimate of the
public reporting burden for this
information collection request under the
final rule is 2900 hours per response.
Public Protection Notification
If a means used to impose an
information collection does not display
a currently valid OMB control number,
the NRC may not conduct or sponsor,
and a person is not required to respond
to the information collection.
VI. Regulatory Analysis
The Commission has prepared a
regulatory analysis for this rulemaking.
The analysis sets forth the objectives of
the rulemaking, the alternatives
considered, and examines the values
and impacts of the alternatives
considered by the Commission. The
alternatives considered in this analysis
include no action, issuance of guidance
only, or rulemaking. The analysis is
available for inspection in the NRC
Public Document Room, 2120 L Street
NW., (Lower Level), Washington, D.C.
VII. Regulatory Flexibility Certification
In accordance with the Regulatory
Flexibility Act of 1980, (5 U.S.C.
605(b)), the Commission certifies that
this rule will not, have a significant
economic impact on a substantial
number of small entities. This rule
affects only the licensing, operation and
decommissioning of nuclear power
plants, nonpower reactors, and
independent spent fuel storage facilities
(including cask certificate holders). The
companies that own these facilities do
not fall within the scope of the
definition of ``small entities'' set forth in
the Regulatory Flexibility Act or the
Small Business Size Standards set out in
regulations issued by the Small
Business Administration at 13 CFR part
121.
VIII. Backfit Analysis
The Commission has evaluated these
rule changes under the backfitting
requirements in §§50.109 and 72.62.
The Commission does not regard the
changes to be backfits as defined in
§§50.109(a)(1) and 72.62(a), as
applicable. Accordingly, a backfit
analysis applicable to these changes has
not been prepared. However, the
Commission has prepared a regulatory
analysis which sets forth the objectives
of the rulemaking changes, the
alternatives that were considered, and
the expected benefits and costs
associated with the rulemaking changes.
The Commission regards this analysis as
providing for a disciplined approach for
evaluating the impacts of the proposed
changes, which satisfies the underlying
purposes of the backfitting requirements
in §§50.109 and 72.62.
Changes to Section 50.59
Section 50.59 defines the
circumstances under which holders of
nuclear power plant operating licenses
may make changes to and conduct tests
or experiments at their facilities without
prior NRC review and approval. In this
rulemaking, new definitions are added
to §50.59 (e.g., the definitions for
``change,'' and ``facility as described in
the final safety analysis report (as
updated)''), and the structure and
language of the rule were modified (e.g.,
the addition of a new applicability
section, and the removal of the term,
``unreviewed safety question''). These
changes constitute clarifications of the
existing rule, and codification of
existing NRC practice and
interpretations of terminology which are
undefined by the current rule.
Clarifications and codification of
existing NRC interpretation and practice
do not constitute a generic backfit
(although the application of the revised
rule may constitute a plant-specific
backfit). The new criteria in
§50.59(c)(2)(i), (ii), (iii), (iv), (v) and (vi)
are being added primarily 4 for the
purpose of providing additional
flexibility to licensees to make changes
and conduct tests without having to
obtain prior NRC review and approval.
Each of these changes constitute
permissive relaxations 5 from the
superseded §50.59(a)(2)(i) and (ii)
criteria. Permissive relaxations are not
considered to be backfits, inasmuch as
a licensee will continue to be in
compliance with the final rule even if it
uses its existing procedures and the
superseded criteria for implementing
§50.59. The new criteria in
§50.59(c)(2)(vii) and (viii) together
constitute replacements for the
superseded §50.59(a)(2)(iii) criterion on
``margin of safety.'' As noted in Section
J, these two criteria together, in place of
a criterion on margin of safety,
explicitly cover those margins that the
Commission believes are important to
address in this evaluation processÐthe
first being the margin that exists in the
limits that are to be met, and the second
being the margin that exists from the
conservatisms included in the methods
used to demonstrate that requirements
are met. The replacement criteria were
thus developed to accomplish two
complementary goals: (1) Defining with
more precision the important safety
margins which should be the focus of a
§50.59 determination, rather than the
problematic term, ``margin of safety as
defined in the basis for any technical
specification;'' and (2) assuring that the
relaxations embodied in the
§50.59(c)(2)(i), (ii), (iii), (iv), (v) and (vi)
criteria will not result in changes
approaching the adequate protection
threshold without prior NRC review and
approval. As such, the new criteria (vii)
and (viii) are fundamentally part of the
overall regulatory scheme in the
revisions to §50.59 which relax and
clarify the thresholds for licensee-
initiated changes and tests requiring
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53612 Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
prior NRC review and approval before
their implementation. In sum, the
Commission has determined that the
changes to §50.59 constitute
clarifications and codifications of
existing practices, or constitute
permissive relaxations from the existing
§50.59 criteria, and therefore do not
constitute backfits as defined in
§50.109(a)(1).
Changes to Part 72
Section 72.48 defines the
circumstances under which a holder of
a ISFSI license may make changes and
conduct tests and experiments,
analogous to the criteria in §50.59. The
change to §72.48 will conform the
criteria for ISFSI and storage cask
changes to that in §50.59. Therefore, as
with the changes to §50.59, the changes
to §72.48 constitute a permissive
relaxation as compared with the existing
criteria in §72.48. Furthermore, there
will be consistency in regulatory
approach in changes to nuclear power
plants and ISFSIs. Such consistency is
appropriate since most ISFSIs are
licensed to nuclear power plant
licensees; there are resource efficiencies
for such licensees using the same
criteria for evaluating changes, tests and
experiments. The change criteria in
§72.48 are also extended by the final
rule to holders of CoCs., which
contributes to regulatory stability and
predictability since known standards
will be utilized in determining whether
a change to a CoC may be made without
prior NRC review and approval. The
existing backfitting provision in §72.62
only apply to licensees and not to CoC
holders. However, even if the backfitting
provisions in §72.62 applied to CoC
holders, the changes in §72.48 would
not be regarded as backfits since the
extension of §72.48 to CoC holders
represents a permissive relaxation. For
similar reasons, the changes in part 72
applicable to CoC holders, which are
necessary to support the extension of
the change criteria in §72.48 to CoC
holders, are not considered to be
backfits under §72.62.
The Commission is deferring
consideration of conforming changes to
the design certifications in part 52,
appendices A and B, which are the
design certifications for the ABWR and
System 80+ designs. The Commission
will conduct a broader rulemaking to
amend part 52, whose purpose will be
to correct typographic errors, clarify
language, and reflect lessons learned as
a result of the ABWR, System 80+, and
AP600 design certification rulemakings.
If conforming changes to appendices A
and B are made, in a future rulemaking,
the Commission regards this rulemaking
amending §50.59 as satisfying the
Commission's obligations under the
backfit rule for any conforming changes
made to part 52, inasmuch as the
backfitting issues associated with the
adoption of the new criteria are being
addressed in this rulemaking.
IX. Small Business Regulatory
Enforcement Fairness Act
In accordance with the Small
Business Regulatory Enforcement
Fairness Act of 1996, the NRC has
determined that this action is not a
major rule and has verified this
determination with the Office of
Information and Regulatory Affairs of
OMB.
X. National Technology Transfer and
Advancement Act
The National Technology Transfer
and Advancement Act of 1995, Pub. L.
104±113, requires that Federal agencies
use technical standards developed by or
adopted by voluntary consensus
standards bodies unless the use of such
a standard is inconsistent with
applicable law or otherwise impractical.
There are no consensus standards that
apply to the change control process
requirements established in this
rulemaking. Thus the provisions of the
Act do not apply to this rulemaking.
XI. Criminal Penalties
For the purposes of section 223 of the
Atomic Energy Act (AEA), the
Commission is issuing this rule to
amend 10 CFR part 50:50.59, : 50.66,
and :50.71; and 10 CFR part 72:72.48, :
72.70, :72.212, and :72.248, under one
or more of sections 161b, 161i, or 161o
of the AEA. Willful violations of the
rule would be subject to criminal
enforcement.
XII. Compatibility of Agreement State
Regulations
Under the ``Policy Statement on
Adequacy and Compatibility of
Agreement State Programs'' approved by
the Commission on June 30, 1997, and
published in the Federal Register (62
FR 46517, September 3, 1997), this rule
is classified as compatibility Category
``NRC.'' Compatibility is not required for
Category ``NRC'' regulations. The NRC
program elements in this category are
those that relate directly to areas of
regulation reserved to the NRC by the
AEA or the provisions of Title 10 of the
Code of Federal Regulations, and
although an Agreement State may not
adopt program elements reserved to
NRC, it may wish to inform its licensees
of certain requirements via a mechanism
that is consistent with the particular
State's administrative procedure laws,
but that does not confer regulatory
authority on the State.
List of Subjects
10 CFR Part 50
Antitrust, Classified information,
Criminal penalties, Fire protection,
Intergovernmental relations, Nuclear
power plants and reactors, Radiation
protection, Reactor siting criteria,
Reporting and record keeping
requirements.
10 CFR Part 72
Criminal penalties, Manpower
training programs, Nuclear materials,
Occupational safety and health,
Reporting and recordkeeping
requirements, Security measures, Spent
fuel.
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 adopting the following
amendments to 10 CFR parts 50 and 72.
PART 50—DOMESTIC LICENSING OF
PRODUCTION AND UTILIZATION
FACILITIES
1. The authority citation for part 50
continues to read as follows:
Authority: Secs. 102, 103, 104, 105, 161,
182, 183, 186, 189, 68 Stat. 936, 937, 938,
948, 953, 954, 955, 956, as amended, sec.
234, 83 Stat. 444, as amended (42 U.S.C.
2132, 2133, 2134, 2135, 2201, 2232, 2233,
2236, 2239, 2282); secs. 201, as amended,
202, 206, 88 Stat. 1242, as amended, 1244,
1246, (42 U.S.C. 5841, 5842, 5846).
Section 50.7 also issued under Pub. L. 95±
601, sec. 10, 92 Stat. 2951, as amended by
Pub. L. 102±486, sec. 2902, 106 Stat. 3123,
(42 U.S.C. 5851). Sections 50.10 also issued
under secs. 101, 185, 68 Stat. 936, 955, as
amended (42 U.S.C. 2131, 2235); sec. 102,
Pub. L. 91±190, 83 Stat. 853 (42 U.S.C. 4332).
Sections 50.13, 50.54(dd), and 50.103 also
issued under sec. 108, 68 Stat. 939, as
amended (42 U.S.C. 2138). Sections 50.23,
50.35, 50.55, and 50.56 also issued under sec.
185, 68 Stat. 955 (42 U.S.C. 2235). Sections
50.33a, 50.55a, and Appendix Q also issued
under sec. 102, Pub. L. 91±190, 83 Stat. 853
(42 U.S.C. 4332). Sections 50.34 and 50.54
also issued under sec. 204, 88 Stat. 1245 (42
U.S.C. 5844). Sections 50.58, 50.91, and
50.92 also issued under Pub. L. 97±415, 96
Stat. 2073 (42 U.S.C. 2239). Sections 50.78
also issued under sec. 122, 68 Stat. 939 (42
U.S.C. 2152). Sections 50.80, 50.81 also
issued under sec. 184, 68 Stat. 954, as
amended (42 U.S.C. 2234). Appendix F also
issued under sec. 187, 66 Stat. 955 (42 U.S.C.
2237).
2. Section 50.59 is revised to read as
follows:
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§50.59 Changes, tests, and experiments.
(a) Definitions for the purposes of this
section:
(1) Change means a modification or
addition to, or removal from, the facility
or procedures that affects a design
function, method of performing or
controlling the function, or an
evaluation that demonstrates that
intended functions will be
accomplished.
(2) Departure from a method of
evaluation described in the FSAR (as
updated) used in establishing the design
bases or in the safety analyses means:
(i) Changing any of the elements of
the method described in the FSAR (as
updated) unless the results of the
analysis are conservative or essentially
the same; or
(ii) Changing from a method described
in the FSAR to another method unless
that method has been approved by NRC
for the intended application.
(3) Facility as described in the final
safety analysis report (as updated)
means:
(i) The structures, systems, and
components (SSC) that are described in
the final safety analysis report (FSAR)
(as updated),
(ii) The design and performance
requirements for such SSCs described in
the FSAR (as updated), and
(iii) The evaluations or methods of
evaluation included in the FSAR (as
updated) for such SSCs which
demonstrate that their intended
function(s) will be accomplished.
(4) Final Safety Analysis Report (as
updated) means the Final Safety
Analysis Report (or Final Hazards
Summary Report) submitted in
accordance with §50.34, as amended
and supplemented, and as updated per
the requirements of §50.71(e) or
§50.71(f), as applicable.
(5) Procedures as described in the
final safety analysis report (as updated)
means those procedures that contain
information described in the FSAR (as
updated) such as how structures,
systems, and components are operated
and controlled (including assumed
operator actions and response times).
(6) Tests or experiments not described
in the final safety analysis report (as
updated) means any activity where any
structure, system, or component is
utilized or controlled in a manner
which is either:
(i) Outside the reference bounds of the
design bases as described in the final
safety analysis report (as updated) or
(ii) Inconsistent with the analyses or
descriptions in the final safety analysis
report (as updated).
(b) Applicability. This section applies
to each holder of a license authorizing
operation of a production or utilization
facility, including the holder of a license
authorizing operation of a nuclear
power reactor that has submitted the
certification of permanent cessation of
operations required under §50.82(a)(1)
or a reactor licensee whose license has
been amended to allow possession but
not operation of the facility.
(c)(1) A licensee may make changes in
the facility as described in the final
safety analysis report (as updated), make
changes in the procedures as described
in the final safety analysis report (as
updated), and conduct tests or
experiments not described in the final
safety analysis report (as updated)
without obtaining a license amendment
pursuant to §50.90 only if:
(i) A change to the technical
specifications incorporated in the
license is not required, and
(ii) The change, test, or experiment
does not meet any of the criteria in
paragraph (c)(2) of this section.
(2) A licensee shall obtain a license
amendment pursuant to §50.90 prior to
implementing a proposed change, test,
or experiment if the change, test, or
experiment would:
(i) Result in more than a minimal
increase in the frequency of occurrence
of an accident previously evaluated in
the final safety analysis report (as
updated);
(ii) Result in more than a minimal
increase in the likelihood of occurrence
of a malfunction of a structure, system,
or component (SSC) important to safety
previously evaluated in the final safety
analysis report (as updated);
(iii) Result in more than a minimal
increase in the consequences of an
accident previously evaluated in the
final safety analysis report (as updated);
(iv) Result in more than a minimal
increase in the consequences of a
malfunction of an SSC important to
safety previously evaluated in the final
safety analysis report (as updated);
(v) Create a possibility for an accident
of a different type than any previously
evaluated in the final safety analysis
report (as updated);
(vi) Create a possibility for a
malfunction of an SSC important to
safety with a different result than any
previously evaluated in the final safety
analysis report (as updated);
(vii) Result in a design basis limit for
a fission product barrier as described in
the FSAR (as updated) being exceeded
or altered; or
(viii) Result in a departure from a
method of evaluation described in the
FSAR (as updated) used in establishing
the design bases or in the safety
analyses.
(3) In implementing this paragraph,
the FSAR (as updated) is considered to
include FSAR changes resulting from
evaluations performed pursuant to this
section and analyses performed
pursuant to §50.90 since submittal of
the last update of the final safety
analysis report pursuant to §50.71 of
this part.
(4) The provisions in this section do
not apply to changes to the facility or
procedures when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(d)(1) The licensee shall maintain
records of changes in the facility, of
changes in procedures, and of tests and
experiments made pursuant to
paragraph (c) of this section. These
records must include a written
evaluation which provides the bases for
the determination that the change, test,
or experiment does not require a license
amendment pursuant to paragraph (c)(2)
of this section.
(2) The licensee shall submit, as
specified in §50.4, a report containing
a brief description of any changes, tests,
and experiments, including a summary
of the evaluation of each. A report must
be submitted at intervals not to exceed
24 months.
(3) The records of changes in the
facility must be maintained until the
termination of a license issued pursuant
to this part or the termination of a
license issued pursuant to 10 CFR part
54, whichever is later. Records of
changes in procedures and records of
tests and experiments must be
maintained for a period of 5 years.
3. In §50.66, paragraph (b),
introductory text, paragraphs (b)(4),
(c)(2), and (c)(3)(iii) are revised to read
as follows:
§50.66 Requirements for thermal
annealing of the reactor pressure vessel.
* * * * *
(b) Thermal Annealing Report. The
Thermal Annealing Report must
include: a Thermal Annealing Operating
Plan; a Requalification Inspection and
Test Program; a Fracture Toughness
Recovery and Reembrittlement Trend
Assurance Program; and an
Identification of Changes Requiring a
License Amendment.
(1) * * *
(4) Identification of Changes
Requiring a License Amendment. Any
changes to the facility as described in
the final safety analysis report (as
updated) which requires a license
amendment pursuant to §50.59(c)(2) of
this part, and any changes to the
Technical Specifications, which are
necessary to either conduct the thermal
annealing or to operate the nuclear
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53614 Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
1Effects of changes includes appropriate
revisions of descriptions in the FSAR such that the
FSAR (as updated) is complete and accurate.
power reactor following the annealing
must be identified. The section shall
demonstrate that the Commission's
requirements continue to be complied
with, and that there is reasonable
assurance of adequate protection to the
public health and safety following the
changes.
(c) * * *
(2) If the thermal annealing was
completed but the annealing was not
performed in accordance with the
Thermal Annealing Operating Plan and
the Requalification Inspection and Test
Program, the licensee shall submit a
summary of lack of compliance with the
Thermal Annealing Operating Plan and
the Requalification Inspection and Test
Program and a justification for
subsequent operation to the Director,
Office of Nuclear Reactor Regulation.
Any changes to the facility as described
in the final safety analysis report (as
updated) which are attributable to the
noncompliances and which require a
license amendment pursuant to
§50.59(c)(2) and any changes to the
Technical Specifications shall also be
identified.
(i) If no changes requiring a license
amendment pursuant to §50.59(c)(2) or
changes to Technical Specifications are
identified, the licensee may restart its
reactor after the requirements of
paragraph (f)(2) of this section have
been met.
(ii) If any changes requiring a license
amendment pursuant to §50.59(c)(2) or
changes to the Technical Specifications
are identified, the licensee may not
restart its reactor until approval is
obtained from the Director, Office of
Nuclear Reactor Regulation and the
requirements of paragraph (f)(2) of this
section have been met.
(3) * * *
(iii) If the partial annealing was not
performed in accordance with the
Thermal Annealing Operating Plan and
the Requalification Inspection and Test
Program, the licensee shall submit a
summary of lack of compliance with the
Thermal Annealing Operating Plan and
the Requalification Inspection and Test
Program and a justification for
subsequent operation to the Director,
Office of Nuclear Reactor Regulation.
Any changes to the facility as described
in the final safety analysis report (as
updated) which are attributable to the
noncompliances and which require a
license amendment pursuant to
§50.59(c)(2) and any changes to the
technical specifications which are
required as a result of the
noncompliances, shall also be
identified.
(A) If no changes requiring a license
amendment pursuant to §50.59(c)(2) or
changes to Technical Specifications are
identified, the licensee may restart its
reactor after the requirements of
paragraph (f)(2) of this section have
been met.
(B) If any changes requiring a license
amendment pursuant to §50.59(c)(2) or
changes to Technical Specifications are
identified, the licensee may not restart
its reactor until approval is obtained
from the Director, Office of Nuclear
Reactor Regulation and the
requirements of paragraph (f)(2) of this
section have been met.
* * * * *
4. In §50.71, paragraph (e),
introductory text is revised to read as
follows:
§50.71 Maintenance of records, making of
reports.
* * * * *
(e) Each person licensed to operate a
nuclear power reactor pursuant to the
provisions of §50.21 or §50.22 of this
part shall update periodically, as
provided in paragraphs (e) (3) and (4) of
this section, the final safety analysis
report (FSAR) originally submitted as
part of the application for the operating
license, to assure that the information
included in the report contains the
latest information developed. This
submittal shall contain all the changes
necessary to reflect information and
analyses submitted to the Commission
by the licensee or prepared by the
licensee pursuant to Commission
requirement since the submittal of the
original FSAR, or as appropriate the last
update to the FSAR under this section.
The submittal shall include the effects1
of: All changes made in the facility or
procedures as described in the FSAR; all
safety analyses and evaluations
performed by the licensee either in
support of approved license
amendments, or in support of
conclusions that changes did not require
a license amendment in accordance
with §50.59(c)(2) of this part; and all
analyses of new safety issues performed
by or on behalf of the licensee at
Commission request. The updated
information shall be appropriately
located within the update to the FSAR.
(1) * * *
* * * * *
5. Section 50.90 is revised to read as
follows:
§50.90 Application for amendment of
license or construction permit.
Whenever a holder of a license or
construction permit desires to amend
the license (including the Technical
Specifications incorporated into the
license) or permit, application for an
amendment must be filed with the
Commission, as specified in §50.4, fully
describing the changes desired, and
following as far as applicable, the form
prescribed for original applications.
PART 72—LICENSING
REQUIREMENTS FOR THE
INDEPENDENT STORAGE OF SPENT
NUCLEAR FUEL AND HIGH-LEVEL
RADIOACTIVE WASTE
6. The authority citation for part 72
continues to read as follows:
Authority: Secs. 51, 53, 57, 62, 63, 65, 69,
81, 161, 182, 183, 184, 186, 187, 189, 68 Stat.
929, 930, 932, 933, 934, 935, 948, 953, 954,
955, as amended, sec. 234, 83 Stat. 444, as
amended (42 U.S.C. 2071, 2073, 2077, 2092,
2093, 2095, 2099, 2111, 2201, 2232, 2233,
2234, 2236, 2237, 2238, 2282); sec. 274, Pub.
L. 86±373, 73 Stat. 688, as amended (42
U.S.C. 2021); sec. 201, as amended, 202, 206,
88 Stat. 1242, as amended, 1244, 1246 (42
U.S.C. 5841, 5842, 5846); Pub. L. 95±601, sec.
10, 92 Stat. 2951 (42 U.S.C. 5851); sec. 102,
Pub. L. 91±190, 83 Stat. 853 (42 U.S.C. 4332);
secs. 131, 132, 133, 135, 137, 141, Pub. L. 97±
425, 96 Stat. 2229, 2230, 2232, 2241, sec. 148,
Pub. L. 100±203, 101 Stat. 1330±235 (42
U.S.C. 10151, 10152, 10153, 10155, 10157,
10161, 10168).
Section 72.44(g) also issued under secs.
142(b) and 148 (c), (d), Pub. L. 100±203, 101
Stat. 1330±232, 1330±236 (42 U.S.C.
10162(b), 10168(c), (d)). Section 72.46 also
issued under sec. 189, 68 Stat. 955 (42 U.S.C.
2239); sec. 134, Pub. L. 97±425, 96 Stat. 2230
(42 U.S.C. 10154). Section 72.96(d) also
issued under sec. 145(g), Pub. L. 100±203,
101 Stat. 1330±235 (42 U.S.C. 10165(g)).
Subpart J also issued under secs. 2(2), 2(15),
2(19), 117(a), 141(h), Pub. L. 97±425, 96 Stat.
2202, 2203, 2204, 2222, 2224 (42 U.S.C.
10101, 10137(a), 10161(h)). Subparts K and L
are also issued under sec. 133, 98 Stat. 2230
(42 U.S.C. 10153) and sec. 218(a), 96 Stat.
2252 (42 U.S.C. 10198).
7. Section 72.3 is amended by revising
the definition for independent spent
fuel storage installation or ISFSI to read
as follows:
§72.3 Definitions.
* * * * *
Independent spent fuel storage
installation or ISFSI means a complex
designed and constructed for the
interim storage of spent nuclear fuel and
other radioactive materials associated
with spent fuel storage. An ISFSI which
is located on the site of another facility
licensed under this part or a facility
licensed under part 50 of this chapter
and which shares common utilities and
services with such a facility or is
physically connected with such other
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53615Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
facility may still be considered
independent.
* * * * *
8. In §72.9, paragraph (b) is revised to
read as follows:
§72.9 Information collection
requirements: OMB approval.
* * * * *
(b) The approved information
collection requirements contained in
this part appear in §§72.7, 72.11, 72.16,
72.19, 72.22 through 72.34, 72.42, 72.44,
72.48 through 72.56, 72.62, 72.70
through 72.82, 72.90, 72.92, 72.94,
72.98, 72.100, 72.102, 72.104, 72.108,
72.120, 72.126, 72.140 through 72.176,
72.180 through 72.186, 72.192, 72.206,
72.212, 72.216, 72.218, 72.230, 72.232,
72.234, 72.236, 72.240, 72.244, and
72.248.
9. In §72.24, paragraph (a) is revised
as follows:
§72.24 Contents of application: Technical
information.
* * * * *
(a) A description and safety
assessment of the site on which the
ISFSI or MRS is to be located, with
appropriate attention to the design bases
for external events. Such assessment
must contain an analysis and evaluation
of the major structures, systems, and
components of the ISFSI or MRS that
bear on the suitability of the site when
the ISFSI or MRS is operated at its
design capacity. If the proposed ISFSI or
MRS is to be located on the site of a
nuclear power plant or other licensed
facility, the potential interactions
between the ISFSI or MRS and such
other facilityÐincluding shared
common utilities and servicesÐmust be
evaluated.
* * * * *
10. Section 72.48 is revised to read as
follows:
§72.48 Changes, tests, and experiments.
(a) Definitions for the purposes of this
section:
(1) Change means a modification or
addition to, or removal from, the facility
or spent fuel storage cask design or
procedures that affects a design
function, method of performing or
controlling the function, or an
evaluation that demonstrates that
intended functions will be
accomplished.
(2) Departure from a method of
evaluation described in the FSAR (as
updated) used in establishing the design
bases or in the safety analyses means:
(i) Changing any of the elements of
the method described in the FSAR (as
updated) unless the results of the
analysis are conservative or essentially
the same; or
(ii) Changing from a method described
in the FSAR to another method unless
that method has been approved by NRC
for the intended application.
(3) Facility means either an
independent spent fuel storage
installation (ISFSI) or a Monitored
Retrievable Storage facility( MRS).
(4) The facility or spent fuel storage
cask design as described in the Final
Safety Analysis Report (FSAR) (as
updated) means:
(i) The structures, systems, and
components (SSC) that are described in
the FSAR (as updated),
(ii) The design and performance
requirements for such SSCs described in
the FSAR (as updated), and
(iii) The evaluations or methods of
evaluation included in the FSAR (as
updated) for such SSCs which
demonstrate that their intended
function(s) will be accomplished.
(5) Final Safety Analysis Report (as
updated) means:
(i) For specific licensees, the Safety
Analysis Report for a facility submitted
and updated in accordance with §72.70;
(ii) For general licensees, the Safety
Analysis Report for a spent fuel storage
cask design, as amended and
supplemented; and
(iii) For certificate holders, the Safety
Analysis Report for a spent fuel storage
cask design submitted and updated in
accordance with §72.248.
(6) Procedures as described in the
Final Safety Analysis Report (as
updated) means those procedures that
contain information described in the
FSAR (as updated) such as how SSCs
are operated and controlled (including
assumed operator actions and response
times).
(7) Tests or experiments not described
in the Final Safety Analysis Report (as
updated) means any activity where any
SSC is utilized or controlled in a
manner which is either:
(i) Outside the reference bounds of the
design bases as described in the FSAR
(as updated) or
(ii) Inconsistent with the analyses or
descriptions in the FSAR (as updated).
(b) This section applies to:
(1) Each holder of a general or specific
license issued under this part, and
(2) Each holder of a Certificate of
Compliance (CoC) issued under this
part.
(c)(1) A licensee or certificate holder
may make changes in the facility or
spent fuel storage cask design as
described in the FSAR (as updated),
make changes in the procedures as
described in the FSAR (as updated), and
conduct tests or experiments not
described in the FSAR (as updated),
without obtaining either:
(i) A license amendment pursuant to
§72.56 (for specific licensees) or
(ii) A CoC amendment submitted by
the certificate holder pursuant to
§72.244 (for general licensees and
certificate holders) if:
(A) A change to the technical
specifications incorporated in the
specific license is not required; or
(B) A change in the terms, conditions,
or specifications incorporated in the
CoC is not required; and
(C) The change, test, or experiment
does not meet any of the criteria in
paragraph (c)(2) of this section.
(2) A specific licensee shall obtain a
license amendment pursuant to §72.56,
a certificate holder shall obtain a CoC
amendment pursuant to §72.244, and a
general licensee shall request that the
certificate holder obtain a CoC
amendment pursuant to §72.244, prior
to implementing a proposed change,
test, or experiment if the change, test, or
experiment would:
(i) Result in more than a minimal
increase in the frequency of occurrence
of an accident previously evaluated in
the FSAR (as updated);
(ii) Result in more than a minimal
increase in the likelihood of occurrence
of a malfunction of a system, structure,
or component (SSC) important to safety
previously evaluated in the FSAR (as
updated);
(iii) Result in more than a minimal
increase in the consequences of an
accident previously evaluated in the
FSAR;
(iv) Result in more than a minimal
increase in the consequences of a
malfunction of an SSC important to
safety previously evaluated in the FSAR
(as updated);
(v) Create a possibility for an accident
of a different type than any previously
evaluated in the FSAR (as updated);
(vi) Create a possibility for a
malfunction of an SSC important to
safety with a different result than any
previously evaluated in the FSAR (as
updated);
(vii) Result in a design basis limit for
a fission product barrier being exceeded
or altered as described in the FSAR (as
updated); or
(viii) Result in a departure from a
method of evaluation described in the
FSAR (as updated) used in establishing
the design bases or in the safety
analyses.
(3) In implementing this paragraph,
the FSAR (as updated) is considered to
include FSAR changes resulting from
evaluations performed pursuant to this
section and analyses performed
pursuant to §72.56 or §72.244 since the
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1Effects of changes includes appropriate
revisions of descriptions in the FSAR such that the
FSAR (as updated) is complete and accurate.
last update of the FSAR pursuant to
§72.70, or §72.248 of this part.
(4) The provisions in this section do
not apply to changes to the facility or
procedures when the applicable
regulations establish more specific
criteria for accomplishing such changes.
(d)(1) The licensee and certificate
holder shall maintain records of changes
in the facility or spent fuel storage cask
design, of changes in procedures, and of
tests and experiments made pursuant to
paragraph (c) of this section. These
records must include a written
evaluation which provides the bases for
the determination that the change, test,
or experiment does not require a license
or CoC amendment pursuant to
paragraph (c)(2) of this section.
(2) The licensee and certificate holder
shall submit, as specified in §72.4, a
report containing a brief description of
any changes, tests, and experiments,
including a summary of the evaluation
of each. A report shall be submitted at
intervals not to exceed 24 months.
(3) The records of changes in the
facility or spent fuel storage cask design
shall be maintained until:
(i) Spent fuel is no longer stored in
the facility or the spent fuel storage cask
design is no longer being used, or
(ii) The Commission terminates the
license or CoC issued pursuant to this
part.
(4) The records of changes in
procedures and of tests and experiments
shall be maintained for a period of 5
years.
(5) The holder of a spent fuel storage
cask design CoC, who permanently
ceases operation, shall provide the
records of changes to the new certificate
holder or to the Commission, as
appropriate, in accordance with
§72.234(d)(3).
(6)(i) A general licensee shall provide
a copy of the record for any changes to
a spent fuel storage cask design to the
applicable certificate holder within 60
days of implementing the change.
(ii) A specific licensee using a spent
fuel storage cask design, approved
pursuant to subpart L of this part, shall
provide a copy of the record for any
changes to a spent fuel storage cask
design to the applicable certificate
holder within 60 days of implementing
the change.
(iii) A certificate holder shall provide
a copy of the record for any changes to
a spent fuel storage cask design to any
general or specific licensee using the
cask design within 60 days of
implementing the change.
11. Section 72.56 is revised to read as
follows:
§72.56 Application for amendment of
license.
Whenever a holder of a specific
license desires to amend the license
(including a change to the license
conditions), an application for an
amendment shall be filed with the
Commission fully describing the
changes desired and the reasons for
such changes, and following as far as
applicable the form prescribed for
original applications.
12. Section 72.70 is revised to read as
follows:
§72.70 Safety analysis report updating.
(a) Each specific licensee for an ISFSI
or MRS shall update periodically, as
provided in paragraphs (b) and (c) of
this section, the final safety analysis
report (FSAR) to assure that the
information included in the report
contains the latest information
developed.
(1) Each licensee shall submit an
original FSAR to the Commission, in
accordance with §72.4, within 90 days
after issuance of the license.
(2) The original FSAR shall be based
on the safety analysis report submitted
with the application and reflect any
changes and applicant commitments
developed during the license approval
and/or hearing process.
(b) Each update shall contain all the
changes necessary to reflect information
and analyses submitted to the
Commission by the licensee or prepared
by the licensee pursuant to Commission
requirement since the submission of the
original FSAR or, as appropriate, the
last update to the FSAR under this
section. The update shall include the
effects1 of:
(1) All changes made in the ISFSI or
MRS or procedures as described in the
FSAR;
(2) All safety analyses and evaluations
performed by the licensee either in
support of approved license
amendments, or in support of
conclusions that changes did not require
a license amendment in accordance
with §72.48;
(3) All final analyses and evaluations
of the design and performance of
structures, systems, and components
that are important to safety taking into
account any pertinent information
developed during final design,
construction, and preoperational testing;
and
(4) All analyses of new safety issues
performed by or on behalf of the
licensee at Commission request. The
information shall be appropriately
located within the updated FSAR.
(c)(1) The update of the FSAR shall be
filed in accordance with §72.4, on a
replacement-page basis;
(2) The update shall include a list that
identifies the current pages of the FSAR
following page replacement;
(3) Each replacement page shall
include both a change indicator for the
area changed, e.g., a bold line vertically
drawn in the margin adjacent to the
portion actually changed, and a page
change identification (date of change or
change number or both);
(4) The update shall include:
(i) A certification by a duly authorized
officer of the licensee that either the
information accurately presents changes
made since the previous submittal, or
that no such changes were made; and
(ii) An identification of changes made
under the provisions of §72.48, but not
previously submitted to the
Commission;
(5) The update shall reflect all
changes implemented up to a maximum
of 6 months prior to the date of filing;
and
(6) Updates shall be filed every 24
months from the date of issuance of the
license.
(d) The updated FSAR shall be
retained by the licensee until the
Commission terminates the license.
13. In §72.80, paragraph (g) is added
to read as follows:
§72.80 Other records and reports.
* * * * *
(g) Each specific licensee shall notify
the Commission, in accordance with
§72.4, of its readiness to begin
operation at least 90 days prior to the
first storage of spent fuel or high-level
waste in an ISFSI or MRS.
14. In §72.86, paragraph (b) is revised
to read as follows:
§72.86 Criminal penalties.
* * * * *
(b) The regulations in this part 72 that
are not issued under sections 161b,
161i, or 161o for the purposes of section
223 are as follows: §§72.1, 72.2, 72.3,
72.4, 72.5, 72.7, 72.8, 72.9, 72.16, 72.18,
72.20, 72.22, 72.24, 72.26, 72.28, 72.32,
72.34, 72.40, 72.46, 72.56, 72.58, 72.60,
72.62, 72.84, 72.86, 72.90, 72.96, 72.108,
72.120, 72.122, 72.124, 72.126, 72.128,
72.130, 72.182, 72.194, 72.200, 72.202,
72.204, 72.206, 72.210, 72.214, 72.220,
72.230, 72.238, 72.240, 72.244, and
72.246.
15. In §72.212, paragraphs (b)(2) and
(b)(4) are revised to read as follows:
§72.212 Conditions of general license
issued under §72.210.
* * * * *
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53617Federal Register /Vol. 64, No. 191/Monday, October 4, 1999/Rules and Regulations
1Effects of changes includes appropriate
revisions of descriptions in the FSAR such that the
FSAR (as updated) is complete and accurate.
(b) * * *
(2)(i) Perform written evaluations,
prior to use, that establish that:
(A) conditions set forth in the
Certificate of Compliance have been
met;
(B) cask storage pads and areas have
been designed to adequately support the
static load of the stored casks; and
(C) the requirements of §72.104 have
been met. A copy of this record shall be
retained until spent fuel is no longer
stored under the general license issued
under §72.210.
(ii) The licensee shall evaluate any
changes to the written evaluations
required by this paragraph using the
requirements of §72.48(c). A copy of
this record shall be retained until spent
fuel is no longer stored under the
general license issued under §72.210.
* * * * *
(4) Prior to use of this general license,
determine whether activities related to
storage of spent fuel under this general
license involve a change in the facility
Technical Specifications or require a
license amendment for the facility
pursuant to §50.59(c)(2) of this chapter.
Results of this determination must be
documented in the evaluation made in
paragraph (b)(2) of this section.
16. Section 72.244 is added to read as
follows:
§72.244 Application for amendment of a
certificate of compliance.
Whenever a certificate holder desires
to amend the CoC (including a change
to the terms, conditions or
specifications of the CoC), an
application for an amendment shall be
filed with the Commission fully
describing the changes desired and the
reasons for such changes, and following
as far as applicable the form prescribed
for original applications.
17. Section 72.246 is added to read as
follows:
§72.246 Issuance of amendment to a
certificate of compliance.
In determining whether an
amendment to a CoC will be issued to
the applicant, the Commission will be
guided by the considerations that
govern the issuance of an initial CoC.
18. Section 72.248 is added to read as
follows:
§72.248 Safety analysis report updating.
(a) Each certificate holder for a spent
fuel storage cask design shall update
periodically, as provided in paragraph
(b) of this section, the final safety
analysis report (FSAR) to assure that the
information included in the report
contains the latest information
developed.
(1) Each certificate holder shall
submit an original FSAR to the
Commission, in accordance with §72.4,
within 90 days after the spent fuel
storage cask design has been approved
pursuant to §72.238.
(2) The original FSAR shall be based
on the safety analysis report submitted
with the application and reflect any
changes and applicant commitments
developed during the cask design
review process. The original FSAR shall
be updated to reflect any changes to
requirements contained in the issued
Certificate of Compliance (CoC).
(b) Each update shall contain all the
changes necessary to reflect information
and analyses submitted to the
Commission by the certificate holder or
prepared by the certificate holder
pursuant to Commission requirement
since the submission of the original
FSAR or, as appropriate, the last update
to the FSAR under this section. The
update shall include the effects1 of:
(1) All changes made in the spent fuel
storage cask design or procedures as
described in the FSAR;
(2) All safety analyses and evaluations
performed by the certificate holder
either in support of approved CoC
amendments, or in support of
conclusions that changes did not require
a CoC amendment in accordance with
§72.48; and
(3) All analyses of new safety issues
performed by or on behalf of the
certificate holder at Commission
request. The information shall be
appropriately located within the
updated FSAR.
(c)(1) The update of the FSAR shall be
filed in accordance with §72.4, on a
replacement-page basis;
(2) The update shall include a list that
identifies the current pages of the FSAR
following page replacement;
(3) Each replacement page shall
include both a change indicator for the
area changed, e.g., a bold line vertically
drawn in the margin adjacent to the
portion actually changed, and a page
change identification (date of change or
change number or both);
(4) The update shall include:
(i) A certification by a duly authorized
officer of the certificate holder that
either the information accurately
presents changes made since the
previous submittal, or that no such
changes were made; and
(ii) An identification of changes made
by the certificate holder under the
provisions of §72.48, but not previously
submitted to the Commission;
(5) The update shall reflect all
changes implemented up to a maximum
of 6 months prior to the date of filing;
(6) Updates shall be filed every 24
months from the date of issuance of the
CoC; and
(7) The certificate holder shall
provide a copy of the updated FSAR to
each general and specific licensee using
its cask design.
(d) The updated FSAR shall be
retained by the certificate holder until
the Commission terminates the
certificate.
(e) A certificate holder who
permanently ceases operation, shall
provide the updated FSAR to the new
certificate holder or to the Commission,
as appropriate, in accordance with
§72.234(d)(3).
Dated at Rockville, Maryland, this 20th day
of September, 1999.
For the Nuclear Regulatory Commission.
Annette Vietti-Cook,
Secretary of the Commission.
[FR Doc. 99±25054 Filed 10±1±99; 8:45 am]
BILLING CODE 7590–01–P
FEDERAL RESERVE SYSTEM
12 CFR Part 204
[Regulation D; Docket No. R–1046]
Reserve Requirements of Depository
Institutions
AGENCY: Board of Governors of the
Federal Reserve System.
ACTION: Final rule.
SUMMARY: The Board is amending
Regulation D, Reserve Requirements of
Depository Institutions, to reflect the
annual indexing of the low reserve
tranche and the reserve requirement
exemption for 2000, and announces the
annual indexing of the deposit reporting
cutoff levels that will be effective
beginning in September 2000. The
amendments decrease the amount of
transaction accounts subject to a reserve
requirement ratio of three percent in
2000, as required by section 19(b)(2)(C)
of the Federal Reserve Act, from $46.5
million to $44.3 million of net
transaction accounts. This adjustment is
known as the low reserve tranche
adjustment. The Board is increasing
from $4.9 million to $5.0 million the
amount of reservable liabilities of each
depository institution that is subject to
a reserve requirement of zero percent in
2000. This action is required by section
19(b)(11)(B) of the Federal Reserve Act,
and the adjustment is known as the
reservable liabilities exemption
adjustment. The Board is also increasing
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