[Federal Register Volume 62, Number 122 (Wednesday, June 25, 1997)]
[Notices]
[Pages 34321-34326]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-16072]
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NUCLEAR REGULATORY COMMISSION
Use of PRA in Plant Specific Reactor Regulatory Activities:
Proposed Regulatory Guides, Standard Review Plan Sections, and
Supporting NUREG
AGENCY: Nuclear Regulatory Commission.
ACTION: Notice of availability.
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SUMMARY: The Nuclear Regulatory Commission has issued for public
comment drafts of four regulatory guides, three Standard Review Plan
Sections, and a NUREG document. These issuances follow Publication of
the Commission's August 16, 1995 (60 FR 42622) Policy statement on the
Use of PRA Methods in Nuclear Regulatory Activities. The NRC has
developed draft guidance for power reactor licensees on acceptable
methods for using probabilistic risk assessment (PRA) information and
insights in support of plant-specific applications to change the
current licensing basis (CLB). The use of such PRA information and
guidance is voluntary. To facilitate comment, the Commission intends to
conduct a workshop during the comment period to explain the draft
documents and answer questions. The exact time, location and agenda
will be announced in a future issue of the Federal Register. Section VI
of this notice provides additional information on the scope, purpose
and topics for discussion at the workshop.
DATES: Comment period expires September 23, 1997. Comments received
after this date will be considered if it is practical to do so, but the
Commission is able to assure consideration only for comments received
on or before this date.
ADDRESSES: Mail written comments to: David L. Meyer, Chief, Rules and
Directives Branch, Office of Administration, Mail Stop T-6D59, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001.
In addition to written comments, please (1) attach a diskette
containing your comments, in either ASCII text or Wordperfect format
(Version 5.1 or 6.1), or (2) submit your comments electronically via
the NRC Electronic Bulletin Board on FedWorld or the NRC's Interactive
Rulemaking Website.
[[Page 34322]]
Deliver comments to 11545 Rockville Pike, Rockville, Maryland,
between 7:30am and 4:15pm, Federal workdays.
Copies of the draft regulatory guides, standard review plan
sections and NUREG are available for inspection and copying for a fee
at the NRC Public Document Room, 2120 L Street NW. (Lower Level),
Washington, DC 20555-0001. A free single copy of these draft documents
to the extent of supply, may be requested by writing to Distribution
Services, Printing, Graphics and Distribution Branch, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, or by Fax to (301)
415-5272. Electronic copies of the draft document are also accessible
on the NRC's Interactive Rulemaking Website through the NRC home page
(http://www.nrc.gov). This site provides the same access as the
FedWorld bulletin board, including the facility to upload comments as
files (any format), if your web browser supports the function.
For more information on the NRC bulletin boards call Mr. Arthur
Davis, Systems Integration and Development Branch, NRC, Washington, DC
20555-0001, telephone (301) 415-5780; e-mail AXD3@nrc.gov. For
information about the Interactive Rulemaking Website, contact Ms. Carol
Gallagher, (301) 415-5905; e-mail [email protected]
The NRC subsystems on FedWorld can be accessed directly by dialing
the toll free number: 1-800-303-9672. Communication software parameters
should be set as follows: parity to none, data bits to 8, and stop bits
to 1 (N,8,1). Using ANSI or VT-100 terminal emulation, the NRC NUREGs
and Reg Guides for Comment subsystem can then be accessed by selecting
the ``Rule Menu'' option from the ``NRC Main Menu.'' For further
information about options available for NRC at FedWorld, consult the
``Help/Information Center'' from the ``NRC Main Menu.'' Users will find
the FedWorld online User's Guides'' particularly helpful. Many NRC
subsystems and databases also have a ``Help/Information Center'' option
that is tailored to the particular subsystem.
The NRC subsystem on FedWorld can also be accessed by a direct dial
phone number for the main FedWorld BBS, 703-321-3339, or by using
Telnet via Internet, fedworld.gov. If using 703-321-3339 to contact
FedWorld, the NRC subsystem will be accessed from the main Fedworld
menu by selecting the ``Regulatory, Government Administration and State
Systems,'' then selecting ``Regulatory, Information Mall.'' At that
point, a menu will be displayed that has an option ``U.S. Nuclear
Regulatory Commission'' that will take you to the NRC Online main menu.
The NRC Online area also can be accessed directly by typing ``/go nrc''
at a FedWorld command line. If you access NRC from FedWorld's main
menu, you may return to FedWorld by selecting the ``Return to
FedWorld'' option from the NRC Online Main Menu. However, if you access
NRC at FedWorld by using NRC's toll-free number, you will have full
access to all NRC systems but you will not have access to the main
Fedworld system.
If you contact FedWorld using Telnet, you will see the NRC area and
menus, including the Rules menu. Although you will be able to download
documents and leave messages, you will not be able to write comments or
upload files (comments). If you contact FedWorld using FTP, all files
can be accessed and downloaded but uploads are not allowed; all you
will see is a list of files without descriptions (normal Gopher look).
An index file listing all files within a subdirectory, with
descriptions, is included. there is a 15-minute time limit for FTP
access.
Although Fedworld can be accessed through the World Wide Web, like
FTP that mode only provides access for downloading files and does not
display the NRC Rules menu.
FOR FURTHER INFORMATION CONTACT: Mark Cunningham, Office of Nuclear
Regulatory Research, MS: T10-E50, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, (301) 415-6189.
SUPPLEMENTARY INFORMATION:
I. Background
On August 16, 1995, (60 FR 42622) the Commission published in the
Federal Register a final policy statement on the Use of Probabilistic
Risk Assessment Methods in Nuclear Regulatory Activities. The policy
statement included the following policy regarding expanded NRC use of
PRA:
1. The use of PRA technology should be increased in all regulatory
matters to the extent supported by the state-of-the-art in PRA methods
and data and in a manner that complements the NRC's deterministic
approach and supports the NRC's traditional defense-in-depth
philosophy.
2. PRA and associated analyses (e.g., sensitivity studies,
uncertainty analyses, and importance measures) should be used in
regulatory matters, where practical within the bounds of the state-of-
the-art, to reduce unnecessary conservatism associated with current
regulatory requirements, regulatory guides, license commitments, and
staff practices. Where appropriate, PRA should be used to support
proposals for additional regulatory requirements in accordance with 10
CFR 50.109 (Backfit Rule). Appropriate procedures for including PRA in
the process for changing regulatory requirements should be developed
and followed. It is, of course, understood that the intent of this
policy is that existing rules and regulations shall be complied with
unless these rules and regulations are revised.
3. PRA evaluations in support of regulatory decisions should be as
realistic as practicable and appropriate supporting data should be
publicly available for review.
4. The Commission's safety goals for nuclear power plants and
subsidiary numerical objectives are to be used with appropriate
consideration of uncertainties in making regulatory judgments on the
need for proposing and backfitting new generic requirements on nuclear
power plant licensees.
It was the Commission's intent that implementation of this policy
statement would improve the regulatory process in three areas:
1. Enhancement of safety decision making by the use of PRA
insights,
2. More efficient use of agency resources, and
3. Reduction in unnecessary burdens on licensees.
In parallel with the development of Commission policy on uses of
risk assessment methods, the NRC developed an agency-wide
implementation plan for application of probabilistic risk assessment
insights within the regulatory process (SECY-95-079). This
implementation plan included tasks to develop Regulatory Guides (RG)
and Standard Review Plans (SRP) in the areas of:
--General guidance,
--Inservice inspection (ISI),
--Inservice testing (IST),
--Technical specification (TS), and
--Graded quality assurance (GQA).
These RGs and SRPs are intended to help implement the Commission's
August 1995 policy on the use of risk information in the regulatory
process and to provide an acceptable approach for power reactor
licensees to prepare and submit and NRC staff to review applications
for proposed plant-specific changes to the current licensing basis that
utilize risk information. Currently, draft RGs/SRPs have been developed
and are ready for comment in the areas of general guidance, IST and TS.
A draft RG for GQA has also been developed and is ready for comment. No
SRP has been developed for GQA, since the NRC staff will utilize its
inspection process
[[Page 34323]]
in the GQA area. In addition, the NRC has prepared draft NUREG-1602,
``Use of PRA in Risk-Informed Applications,'' to provide reference
information for licensees and NRC staff and it is also ready for public
comment. Each of these documents is discussed in more detail below.
II. An Overview of Draft RGs, SRPs, and NUREG-1602
The specific documents available for comment are:
Draft regulatory guide DG 1061, ``An Approach for Using
Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-
Specific Changes to the Current Licensing Basis,'' and its companion
SRP, Chapter 19,
Draft regulatory guide DG-1062 ``An Approach for Plant-
Specific, Risk-Informed, Decision Making: Inservice Testing'' and its
companion SRP, Chapter 3.9.7,
Draft regulatory guide DG-1064, ``An Approach for Plant-
Specific, Risk-Informed Decision Making: Graded Quality Assurance,''
Draft regulatory guide DG-1065, ``An Approach for Plant-
Specific, Risk-Informed Decision Making: Technical Specifications'' and
its companion SRP, Chapter 16.1, and
Draft NUREG-1602, ``Use of PRA in Risk-Informed
Applications.''
The purpose of the RGs and SRPs is to provide guidance to power
reactor licensees and NRC staff reviewers on an acceptable approach for
utilizing risk information to support requests for changes in a plant's
CLB. The purpose of NUREG-1602 is to provide reference information
useful in making decisions on the scope and attributes of PRA. The RGs
describe an alternate means by which licensees can propose plant-
specific CLB changes under 10 CFR Part 50. Adopting the approach of
these RGs is voluntary. Licensees submitting applications for changes
to their CLB may use this approach or an alternative equivalent
approach. To encourage the use of risk information in such
applications, the staff intends to give priority to applications for
burden reduction that use risk information as a supplement to
traditional engineering analyses, consistent with the intent of the
Commission's policy. All applications that improve safety will continue
to receive high priority.
The general RG/SRP have been developed to provide an overall
framework and guidance that is applicable to any proposed CLB change
where risk insights are used to support the change. The application-
specific RGs/SRPs (i.e., IST, TS, GQA) build upon and supplement the
general guidance for proposed CLB changes in their respective technical
areas. Each application-specific RG/SRP references the general RG/SRP,
states that the general guidance is applicable and provides additional
guidance specific to the technical area being addressed.
The guidance provided in these documents is designed to encourage
licensees to use risk information by defining an acceptable framework
for the use of risk information on a plant-specific basis, and by
promoting consistency in PRA applications. It is expected that the
long-term use of risk information in plant-specific licensing actions
will result in improved safety by focusing attention on the more risk
significant aspects of plant design and operation. The draft guidance
provides flexibility to licensees by allowing them to define the scope
of the analysis required to support their proposed change and to
perform appropriate analysis to justify proposed changes to the plant's
CLB.
In conjunction with developing these RGs and SRPs, the staff has
also been working with several licensees on pilot applications of risk
informed regulation in the technical areas listed above. The knowledge
gained to date in interacting with licensees on these pilot
applications has been used to help define the content and guidance
contained in these RGs/SRPs. Additional interactions are expected over
the next several months as work on these pilot applications continues
and licensees and other interested persons have an opportunity to
review the draft RGs/SRPs. The results of these additional interactions
will be factored into the final RGs/SRPs.
III. Policy Issues
On May 15, 1996, the Commission requested the staff to identify and
recommend resolution of the following four policy issues associated
with risk-informed changes to a plant's CLB:
The role of performance-based regulation,
Plant-specific application of safety goals,
Risk neutral vs. increases in risk,
implementation of changes to risk-informed IST and ISI
requirements.
On January 22, 1997, the Commission provided the following guidance
on these issues:
A. The Role of Performance-Based Regulation in the PRA Implementation
Plan
The Commission instructed the staff to include, where practical,
performance-based strategies in the implementation of the risk-informed
regulatory process. Furthermore, the Commission indicated that
application of performance-based approaches should not be limited to
risk-informed initiatives and that performance-based initiatives that
do not explicitly reference criteria derived from PRA insights should
not be excluded from consideration. The Commission also instructed the
staff to include in the PRA Implementation Plan, or in a separate plan,
how these performance-based initiatives will be phased into the overall
regulatory improvement and oversight program and to solicit input from
industry on (or develop on its own) additional performance-based
objectives which are not amenable to probabilistic risk analysis but
could be ranked according to, for example, a relative hazards analysis,
and phase in these initiatives.
B. Plant-Specific Application of Safety Goals
The Safety Goals policy statement, issued by the Commission in
1986, established two qualitative safety goals to help ensure that
nuclear power plant operations do not significantly increase risk to
individuals or to the society. The policy statement also defined two
Quantitative Health Objectives (QHO) for use ``in determining
achievement of the qualitative goals.'' Subsequently, the Commission
approved for use two subsidiary objectives derived from the Safety Goal
QHOs, one on core-damage frequency and one on containment performance,
for use in assessing reactor designs for generic actions. The
Commission approved the Safety Goals for use in generic actions with
the intent that they would define ``how safe is safe enough'' in
deciding how far to go when proposing safety enhancements.
The staff has considered the need for risk guidelines to support
regulatory decision-making in plant-specific circumstances, recognizing
that the use of risk information remains complementary to traditional
engineering analysis and judgment. Specifically, the staff recommended
the development of guidelines for plant-specific applications, derived
from the Commission's current Safety Goals and/or subsidiary objectives
and requested Commission approval.
The Commission tentatively approved the plant-specific application
of safety goals and/or their subsidiary objectives.
C. Risk Neutral vs. Increases in Risk
This policy issue is related to whether to allow small increases in
calculated plant risk in approving a change to the CLB.
[[Page 34324]]
The Commission approved small increases in risk under certain
conditions, for proposed changes to a plant's CLB. In giving this
approval the Commission noted that the terms ``small'' and ``under
certain conditions'' require more precise definition. The staff was
requested to provide a sound rationale for judging small increases and
provide for explicit consideration of uncertainties. Criteria for
judging small increases in risk should be considered in the context of
maintaining reasonable assurance that there is no undue risk to public
health and safety.
Moreover, the Commission asked the staff that, in its development
of risk-informed guidance and review of applications regarding risk-
informed initiatives, to evaluate all safety impacts of proposed
changes in an integrated manner including the use of risk insights to
identify areas where requirements should be increased or improvements
could/should be implemented.
D. Implementation of Changes to Risk-Informed IST and ISI Requirements
This policy issue is related to identifying a means for
implementing risk-informed inservice inspection and testing programs
until rulemaking is complete. The alternatives are to treat proposed
changes as exceptions to 10 CFR 50.55(a) or to treat them as authorized
alternatives under the current rule. The Commission approved risk
informed ISI and IST changes as authorized alternatives under 10 CFR
50.55a(a)(3)(i) to approve the pilot plant applications, provided
appropriate findings can be made. In addition, the Commission
instructed the staff that in cases where the findings necessary to
approve the alternative cannot be made, then the use of exemptions
should be considered.
IV. Structure, Guidelines and Rationale for RGs/SRPs
The approach described in each of the RGs/SRPs has four basic
steps. These are:
--Define the proposed change;
--Perform an integrated engineering analysis (which includes both
traditional engineering and risk analysis) and use of an integrated
decision process;
--Monitoring and feedback to verify assumptions and analysis; and
--Document and submit proposed change.
Five fundamental safety principles are described which should be
met in each application for a change in the CLB. These principles are:
--The proposed change meets the current regulation. This principle
applies unless the proposed change is explicitly related to a requested
exemption or rule change (i.e., a 50.12 ``specific exemption'' or a
2.802 ``petition for rulemaking'');
--Defense-in-depth is maintained;
--Sufficient safety margins are maintained;
--Proposed increases in risk, and their cumulative effect, are small
and do not cause the NRC Safety Goals to be exceeded;
--Performance-based implementation and monitoring strategies are
proposed that address uncertainties in analysis models and data and
provide for timely feedback and corrective action.
These principles represent fundamental safety practices that the
staff believes must be retained in any change to a plant's CLB to
maintain reasonable assurance that there is no undue risk to public
health and safety. Each of these principles is to be considered in the
integrated engineering analysis and decision-making process.
The guidelines for assessing risk proposed in the RGs/SRPs are
derived from the Commission's Safety Goal Quantitative Health
Objectives (QHOs). Specifically, the subsidiary objectives of Core
Damage Frequency (CDF) and Large Early Release Frequency (LERF) are
used as the measures of risk against which changes in the CLB will be
assessed, in lieu of the QHOs themselves, which require level 3 PRA
information (offsite health effects). These were chosen to simplify the
scope of PRA analysis needed, to avoid the large uncertainties
associated with level 3 PRA analysis, and to be consistent with
previous Commission direction to decouple siting from plant design.
The values used in the RGs/SRPs as guidelines for CDF and LERF were
selected to be consistent with the Safety Goal QHOs and previous
Commission guidance. Specifically, a CDF value of 10-4/RY is
proposed as the guideline where further increases in CDF would not be
acceptable (i.e., plants with CDF 10-4/RY would
be expected to propose changes that result in CDF decreases or are
neutral). The CDF value of 10-4/RY is the value endorsed by
the Commission in a Staff Requirements Memorandum dated June 15, 1990,
as a benchmark objective for accident prevention. For plants with CDFs
<>-4/RY, guidelines are proposed on changes in CDF
(CDF) that ensure increases in risk from CLB changes are made
in small steps and that increased NRC management attention is provided
for proposed changes that approach the guidelines (i.e., CDFs in the
range 10-5/RY-10-4/RY and
CDF>10-6/RY). The use of small steps is consistent
with a measured approach (allowing time for monitoring, feedback and
corrective action) and the values chosen for CDF are
consistent with the Commission's Regulatory Analysis Guidelines (NUREG/
BR-0058, Rev. 2).
The guidelines on LERF are derived from the Commission's Safety
Goal QHO for early fatality risk. A LERF value of 10-5/RY is
proposed as the guideline where further increases in LERF would not be
acceptable (i.e., plants with a LERF 10-5/RY
would be expected to propose changes that result in LERF decreases or
are neutral). Similar to CDF, a range is proposed where increased NRC
management attention is required if LERF approaches the guideline
(i.e., LERF in the range of 10-6/RY to 10-5/RY).
The value of 10-5/RY for the LERF guideline corresponds to
that value, estimated from existing PRA results, necessary to ensure
that the early-fatality QHO would be met without undue conservatism. In
effect, the guideline value for LERF is a surrogate for the
Commission's QHO on early fatality risk. Guidelines for changes in LERF
(LERF) are used that limit increases in risk to small values
(i.e., LERF <>-6/RY) to ensure that increases are
made in small increments, are consistent with the Regulatory Analysis
Guidelines and, similar to CDF, require increased management
attention when they approach the guideline value (i.e., LERF
in the range of 10-7/RY to 10-6/RY).
The CDF/CDF and LERF/LERF guidelines are intended
for comparison with a full-scope PRA (i.e., full power, low power and
shutdown conditions and internal and external events). It is expected
that the cumulative impact of previous CLB changes will also be
reflected in the PRA. However, it is recognized that less than full-
scope PRA analysis will likely be acceptable for many proposed CLB
changes and the RG/SRP guidance is intended to allow licensees
flexibility to do analyses appropriate for their proposed change and to
allow the use of qualitative factors in the decision process. In
addition, mean values of CDF and LERF are to be compared against the
guidelines. However, when a proposed change is closer to the
guidelines, a more comprehensive uncertainty and sensitivity analysis
is expected that includes the consideration of qualitative factors.
Only general guidelines on uncertainty/sensitivity analyses are
included in the RGs/SRPs to allow
[[Page 34325]]
licensees flexibility to provide analyses appropriate for their
specific application.
Monitoring and feedback strategies are to be utilized in
implementing the proposed CLB change to help verify assumptions and
analysis and to allow for corrective action should performance be less
than assumed in the analysis. In addition, NRC expects licensees to
identify how and where their proposed changes will be documented as
part of the plant's CLB. This should include documentation that clearly
establishes the basis for the change, ensures that commitments are
known and provides sufficient documentation to allow inspection and
enforcement, if appropriate. Related to the above, since these RGs/SRPs
allow the use of risk information and monitoring programs to support
CLB changes associated with safety related systems, structures and
components (SSCs), it is reasonable to expect that the quality of these
analyses and monitoring programs should be consistent with the quality
of other analyses and activities associated with safety related SSCs
(i.e., 10 CFR part 50, Appendix B, ``Quality Assurance Criteria for
Nuclear Power Plants and Fuel Reprocessing Plants''). Accordingly, DG-
1061 includes guidance regarding quality assurance, including that
associated with the PRA,that ensures the pertinent requirements of 10
CFR part 50, Appendix B are met. In addition, the draft RGs/SRPs use
the definition of CLB that is currently in 10 CFR part 54 ``License
Renewal.'' Although not officially incorporated in 10 CFR part 50, this
definition is considered appropriate for use in these RGs/SRPs.
As mentioned above, the draft guidance encourages licensees to
utilize risk insights to improve safety, as well as to propose
reductions of unnecessary burdens. The Commission's Safety Goals, their
subsidiary objectives and Regulatory Analysis Guidelines have been used
to derive guidelines for judging the acceptability of any calculated
risk increases associated with the proposed CLB change. In this regard,
a measured approach to reviewing and accepting changes to CLBs that
increase risk has been taken. Specifically, the guidelines used
correspond to small calculated increases in risk. In theory, one could
construct an even more generous regulatory framework for consideration
of those risk-informed changes which may have the effect of increasing
risk to the public. Such a framework would include, of course,
assurance of continued adequate protection (that level of protection of
the public health and safety which must be reasonably assured
regardless of economic cost), but it could also include provision for
possible elimination of all measures not needed for adequate protection
which either do not contribute to a substantial reduction in overall
risk or result in continuing costs which are not justified by the
safety benefits. However, a more restrictive practice has been used
which would permit only small increases in risk, and then only when it
is reasonably assured, among other things, that sufficient defense in
depth and safety margins are maintained. This practice is used because
of the uncertainties in PRA and to account for the fact that safety
issues continue to emerge regarding design, construction, and
operational matters notwithstanding the maturity of the nuclear power
industry. In addition, limiting risk increases to small values is
considered prudent until such time as experience is obtained with the
methods and applications discussed in the RGs/SRPs.
V. Comments
The staff is soliciting comments related to the guidance described
in the draft RGs, SRPs and NUREG-1602. Comments submitted by the
readers of this FRN will help ensure that these draft documents have
appropriate scope, depth, quality, and effectiveness. Alternative
views, concerns, clarifications, and corrections expressed in public
comments will be considered in developing the final documents.
VI. Workshop
The Commission intends to conduct a workshop to discuss and explain
the material contained in the draft guides, SRPs and NUREG-1602, and to
answer questions and receive comments and feedback on the proposed
documents. The purpose of the workshop is to facilitate the comment
process. In the workshop the staff will describe each document, its
basis and solicit comment and feedback on their completeness,
correctness and usefulness. Since these documents cover a wide range of
technical areas, many topics will be discussed. Listed below are topics
on which discussion and feedback are sought at the workshop:
(1) Overall Approach
(A) Is it appropriate to apply the Commission's Safety Goals and
their subsidiary objectives on a plant specific basis?
(B) Is it appropriate to allow, under certain conditions, changes
to a plant's CLB that increase CDF and/or LERF?
(C) Is the level of detail in the guidance contained in the
proposed Regulatory Guides and SRPs clear and sufficient, or is more
detailed guidance necessary? What level of detail is needed?
(D) Are the four elements of the risk-informed process described in
the Reg Guides and SRPs clear and sufficient?
(E) Is the guidance on the treatment of uncertainties clear and
sufficient, or is additional guidance necessary? What additional
guidance is needed?
(F) Is guidance on the acceptability and treatment of temporary
changes in the CLB (i.e., temporary changes in risk) needed? If so,
what guidance and acceptance guidelines should be included? Should the
guidance be different for full-power operation vs a shutdown condition?
(G) Is it appropriate to use the definition of ``current licensing
basis'' included in 10 CFR 54 ``License Renewal,'' in these RGs/SRPs?
What other definition would be more appropriate?
(H) Should licensees be required to submit risk information in
support of proposed changes to their CLB?
(I) Are the guidelines for quality described in DG-1061 sufficient
to ensure appropriate quality in those activities that support proposed
changes to the CLB for safety related systems, structures and
components? Are the appropriate provisions from 10 CFR 50, Appendix B,
``Quality Assurance Criteria for Nuclear Power Plants and Fuel
Reprocessing Plants'' applied to the PRA?
(J) Should a licensee's PRA be required to be included in the NRC's
docket file and updated as necessary to reflect previous changes and
recent operating experience?
(K) What other areas, besides graded QA, Tech Specs, IST and ISI
could this process and these guidelines be applied to?
(2) Engineering Evaluation
(A) Are the proposed safety principles clear and sufficient? What
should be clarified and/or added?
(B) Is sufficient guidance provided regarding the intent, scope,
and level of detail requested in the submittal with respect to the
evaluation of the safety principles? What should be added? For example:
1. Should there be different guidance on defense-in-depth for those
items analyzed in the PRA versus those not analyzed? What should the
differences be?
2. Should there be quantitative guidelines for determining the
sufficiency of defense-in-depth and safety margins?
[[Page 34326]]
(C) Is the guidance associated with the probabilistic analysis
sufficient? For example:
1. Is additional guidance on the use of qualitative risk
evaluations necessary? What additional guidance would be appropriate?
2. Are the proposed acceptance guidelines for CDF and LERF and
changes in CDF and LERF appropriate? Are they too restrictive or too
liberal? What guidelines would be more appropriate?
3. Is more specific or less detailed guidance needed on comparison
of PRA results with the CDF and LERF and the CDF and
LERF guidelines?
4. Should there be additional guidance on the number of proposed
risk increases which can be submitted in any given year?
5. Should there be separate LERF guidelines for PWRs and BWRs? What
should they be?
6. Should there be separate LERF guidelines for shutdown
conditions/external events? What should they be?
7. Should there be a guideline on long term release frequency to
supplement LERF? What should it be based upon?
8. Is the guidance in Appendix B of DG-1061 for estimating LERF
sufficient? What else is needed? (It should be noted that the staff
intends to expand this guidance to cover shutdown conditions and
external events).
9. Should there be acceptance guidelines for the use of PRA level 3
(segment of PRA that includes estimation of consequences/health effects
and risk to the public) information? What guidelines would be
appropriate?
10. Should the acceptance guidelines specify a confidence level
that the PRA results should meet when being compared to the risk
guidelines? What is an appropriate confidence level?
11. Should a confidence level or uncertainty level be used to
define the ``management attention'' region in, lieu of a CDF and LERF
range?
(3) Performance Monitoring and Feedback
(A) Should the use of performance monitoring be more widely applied
in regulation and regulatory practice, or is it sufficient to implement
it through the elements described in the proposed Regulatory Guides?
(B) Is performance monitoring and feedback an appropriate element
of the risk-informed process? Should it be used to a greater or lesser
degree?
(C) Is the guidance on performance monitoring and feedback clear
and sufficient? What should be improved?
(4) Graded Quality Assurance Regulatory Guide (DG-1064)
(A) Is the approach for determining the safety-significance of
plant SSCs appropriate? Is it sufficient to identify high and low
safety significant categories? Is the amount of risk analysis overly
burdensome relative to the potential benefits?
(B) Is the guidance in the proposed regulatory guide regarding the
content of QA programs for low safety significant SSCs appropriate?
What additional guidelines are needed, and/or what portions of the
proposed guidelines should be deleted?
(C) Are there any quantitative data that can be used to assess the
risk impact (i.e., CDF or LERF) of reducing QA controls on equipment
performance?
(D) Is the proposed scope of graded QA, that includes safety-
related and other important plant equipment as covered by the
Maintenance Rule, appropriate?
(E) Is the guidance on equipment-performance-monitoring strategies
sufficient?
(F) Is the guidance sufficient regarding the QA controls for
safety-significant, but non-safety-related, equipment that should be
included in the licensee's QA program? What guidance should be
included?
(G) Should the guidance allow for further removal of QA
requirements? In what areas should this be done and what guidance would
be appropriate? For example, is it appropriate for a graded QA program
to eliminate all requirements associated with some of the 18 criteria
specified in 10 CFR part 50, Appendix B?
(5) Technical Specifications Regulatory Guide (DG-1065) and SRP
(A) Are the proposed acceptance guidelines on incremental
conditional core damage probability and incremental conditional large
early release probability from a single AOT change (5E-07 and 5E-08,
respectively) appropriate?
(B) Should there be a guideline on maximum conditional CDF/LERF
during an AOT? What should it be?
(6) Inservice Testing Regulatory Guide (DG-1062) and SRP
(A) PRA models of component unavailability typically use a
parameter lambda () to characterize the component's failure
rate, and this parameter is often considered to be a constant value. Is
the assumption of constant value for realistic? What
different values might be more realistic and what evidence (data)
supports the alternate values?
(B) Is it appropriate, as part of a risk-informed program, to
require licensees to look outside the ASME code boundary and identify
candidate components for testing and then apply ASME criteria to the
conduct of those tests? What is a reasonable way to deal with
relatively high-risk components that are not part of a currently
prescribed IST program?
(C) Is it appropriate to use the ``other acceptable methods''
provision of 10 CFR 50.55a to implement changes to the CLB?
(7) NUREG-1602
(A) Draft NUREG-1602 provides reference material on the scope and
quality of a PRA. Is the information in draft NUREG-1602 complete and
correct? Is it useful as reference material in making assessments on an
application specific basis on the scope and quality of the risk
assessment to support that particular application? How could it be
improved? For example, should it specify acceptable PRA methods?
(B) Would draft NUREG-1602 be useful as a starting point to develop
a national consensus standard on PRA? What would be needed?
(C) Is a national consensus standard on PRA needed or desirable?
VII. Paperwork Reduction Act Statement
These draft regulatory guides contain information collections that
are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et
seq.). These regulatory guides will be submitted to the Office of
Management and Budget for review and approval of the information
collections before the final guides are published.
VIII. Public Protection Notification
The NRC may not conduct or sponsor, and a person is not required to
respond to, an information collection unless it displays a currently
valid OMB control number.
Dated at Rockville, Maryland, this 13th day of June, 1997.
For the Nuclear Regulatory Commission.
John C. Hoyle,
Secretary of the Commission.
[FR Doc. 97-16072 Filed 6-18-97; 8:45 am]
BILLING CODE 7590-01-P
