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U. S. NUCLEAR REGULATORY COMMISSION |
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No. 96-114 FOR IMMEDIATE RELEASE
(Tuesday, August 20, 1996)
NOTE TO EDITORS:
The Nuclear Regulatory Commission has received the three attached reports from its
Advisory Committee on Reactor Safeguards. The reports, in the form of letters, provide
comments on:
--The application of risk-informed, performance-based regulation of nuclear power plants.
--Policy and key technical issues pertaining to Westinghouse Electric's AP600
standardized reactor design.
--Design changes proposed by ABB-Combustion Engineering relating to certification of
its System 80+ reactor.
#
Attachments:
As stated
.
August 15, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: RISK-INFORMED, PERFORMANCE-BASED REGULATION AND RELATED
MATTERS
During the 433rd meeting of the Advisory Committee on Reactor Safeguards, August 8-10, 1996,
we discussed the issues identified in the Staff Requirements Memorandum dated May 15, 1996. We
also discussed the pilot applications for risk-informed, performance-based regulation. Our
Subcommittee on Probabilistic Risk Assessment (PRA) met with representatives of the NRC staff and
the nuclear industry on July 18 and August 7, 1996. We also had the benefit of the documents
referenced.
The staff presentations dealt only with the development of guidelines from the Commission's safety
goals to be used as an element of the evaluation of licensee-initiated changes to licensing
commitments. All of our comments address the application of risk-informed regulation in that
context. At a later time, we will discuss the larger question of the application of the safety goals on
a plant-specific basis.
CONCLUSIONS
Issue 1: Should the Commission's safety goals and subsidiary objectives be referenced or used to
derive guidelines for plant-specific applications and, if so, how?
We believe the safety goals and subsidiary objectives can and should be used to derive guidelines for
plant-specific applications. It is, however, impractical to rely exclusively on the Quantitative Health
Objectives (QHOs) for routine use on an individual plant basis. Criteria based on core damage
frequency (CDF) and large, early release frequency (LERF) focus more sharply on safety issues and
can provide assurance that the QHOs are met.
They should be used in developing detailed guidelines.
.Issue 2: How are uncertainties to be accounted for?
This is a difficult issue. There are models and formal methods to account explicitly for a large number
of uncertainties. However, other uncertainties are unquantifiable. The staff proposes to explore a
number of options, such as establishing margins in the acceptance guidelines, placing more
importance on defense-in-depth, and others, to deal with such uncertainties. Such approaches seem
appropriate, although much work remains to be done.
Issue 3: Should requested changes to the current licensing basis be risk-neutral or should increases
be permitted?
We agree with the staff and industry that increases in risk should be permitted in some situations.
Acceptance guidelines expressed in terms of the proposed change in risk and the current risk
estimates should have three regions: a region in which some increase in risk is acceptable, one in
which it is unacceptable, and one in which further analysis and evaluation would be required.
Issue 4: How should performance-based regulation be implemented in the context of risk-informed
regulation?
We agree with the staff that, where practical, performance-based strategies should be included in the
implementation and monitoring step of the risk-informed decision-making process. The pilot
programs may provide an opportunity for a more concrete definition and development of
performance-based strategies.
DISCUSSION
Issue 1
Even though a CDF could be derived from the QHOs that could be greater than 10-3 per reactor-year,
the current subsidiary goal of 10-4 per reactor-year should be maintained and should be stated as a
fundamental safety goal, along with the QHO. Accident sequences that have a high probability of
leading to severe consequences could be controlled by the QHOs, but a more workable measure
would be a subsidiary goal on the LERF. The definition of the latter needs to be improved. Whether
the LERF should be a fixed value or derived from the QHOs, which would allow the LERF goal to
include site-specific characteristics, needs to be investigated.
We recommend that the staff develop guidance for handling situations in which high values of the
CDF occur for short periods of time (for example, 10-2 per reactor-year for a day).
.Issue 2
In accounting for uncertainties, it is important to distinguish between those plant characteristics or
phenomena that are modeled in the PRA and those that are not modeled (e.g., the actual layout of
components and organizational factors). For those that are modeled, parameter and model
uncertainties should be explicitly quantified and propagated through the PRA. The resulting
distributions should be an input to the decision-making process along with other qualitative input.
Mean values of distributions should, in general, be used for comparison with goals or criteria,
although the sensitivity of the mean value to the high tail of a distribution should not be overlooked.
For very broad distributions, such as those that typically result when significant model uncertainty is
present, reliance on the mean values may not be appropriate and a more detailed investigation of the
reasons for this large uncertainty should be undertaken. This could possibly lead to decisions to
conduct additional research or to take other measures.
Accounting for uncertainty in the case of plant characteristics or phenomena that are not currently
modeled at all is much more difficult. The staff proposes to explore a number of options, such as
establishing margins in the acceptance guidelines, placing more importance on defense-in-depth, and
others. We agree and encourage the staff to actively pursue the resolution of this issue.
Issue 3
The concept of a "three-region" approach is consistent with the Electric Power Research Institute's
PSA Applications Guide (PSAAG), although the boundaries of the regions used in the PSAAG are
not necessarily the ones that the staff will adopt.
The staff has raised the issue of how "packaged" requests are to be handled. Packaging is the process
by which risk trade-offs can be accomplished. It is a significant benefit of risk-informed regulation.
We believe that it is the overall impact on plant risk that is important, and related changes should be
handled as a package. Such changes should be consistent with the current philosophy of risk
management; i.e., that the "bottom-line" numbers should not be the only input to the decision-making
process, and other concepts such as defense-in-depth must be maintained.
.
We will continue to monitor the progress of the staff on these issues.
Sincerely,
/s/
T. S. Kress
Chairman, ACRS
References:
1. Staff Requirements Memorandum dated May 15, 1996, from John C. Hoyle, Secretary, NRC,
to James M. Taylor, Executive Director for Operations, NRC, regarding Briefing on PRA
Implementation Plan on April 4, 1996
2. Memorandum dated June 20, 1996, from James M. Taylor, Executive Director for
Operations, NRC, to the Commission, Subject: Status Update of the Agency-Wide
Implementation Plan for Probabilistic Risk Assessment (PRA) (from March 1, 1996 to May
31, 1996)
3. Electric Power Research Institute, EPRI TR-105396, Final Report dated August 1995, "PSA
Applications Guide"
.
August 15, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: SECY-96-128, "POLICY AND KEY TECHNICAL ISSUES PERTAINING TO
THE WESTINGHOUSE AP600 STANDARDIZED PASSIVE REACTOR
DESIGN"
During the 433rd meeting of the Advisory Committee on Reactor Safeguards, August 8-10, 1996,
we reviewed the subject document. Our Subcommittee on Westinghouse Standard Plant Designs met
on July 19, 1996 to review this matter. During this review, we had the benefit of discussions with
representatives of the staff and of the Westinghouse Electric Corporation. We also had the benefit
of the documents referenced.
Conclusion
We endorse the positions recommended by the staff in addressing the following three policy issues
pertaining to the Westinghouse AP600 standardized passive reactor design.
Policy Issues
. Prevention and Mitigation of Severe Accidents
The staff is seeking Commission approval to consider the use of non-safety systems in the
AP600 design to address the uncertainties associated with the passive fission product removal
mechanisms for design-basis analysis and for balance between prevention and mitigation of
severe accidents. Westinghouse has no objection to the staff's crediting of non-safety
equipment that is already a part of the AP600 design, but objects to a requirement for adding
a non-safety-grade containment spray system.
The applicant's submittals provide some support for demonstrating fission product removal
using only passive removal mechanisms. Nonetheless, we are persuaded by the staff position
that systems beyond the passive removal mechanisms should be evaluated to provide greater
confidence in the performance of the plant design in mitigating design-basis and severe
accidents. We recommend Commission approval.
. External Reactor Vessel Cooling
The staff is seeking Commission approval for requiring that the applicant provide limited
analytical evaluation of postulated ex-vessel phenomena, notwithstanding that the AP600
design is intended to prevent reactor vessel melt-through. We recommend Commission
approval.
. Post-72-hour Actions
The staff is seeking Commission approval for requiring that the AP600 design be capable of
sustaining all design-basis events with onsite equipment and supplies for the long term. We
recommend Commission approval.
Technical Issues
The staff added spent fuel pool cooling to its list of technical issues being tracked in the
review. At present, the applicant will be required to provide additional onsite capability to
remove decay heat from the spent fuel pool over an extended period of time. We believe this
requirement may be found unnecessary after considering the low risk associated with the
current design.
Dr. Dana A. Powers did not participate in the Committee's deliberations regarding the severe accident
source term. Dr. T. S. Kress did not participate in the Committee's deliberations regarding external
reactor vessel cooling.
Sincerely,
/s/
T. S. Kress
Chairman, ACRS
References:
1. U. S. Nuclear Regulatory Commission, SECY-96-128, dated June 12, 1996, from James M.
Taylor, Executive Director for Operations, NRC, to the Commissioners, Subject: Policy and
Key Technical Issues Pertaining to the Westinghouse AP600 Standardized Passive Reactor
Design
2. Letter dated June 15, 1995, from T.S. Kress, Chairman, ACRS, to James M. Taylor,
Executive Director for Operations, NRC, Subject: Proposed Commission Paper on Staff
Positions on Technical Issues Pertaining to the Westinghouse AP600 Standardized Passive
Reactor Design
3. Letter dated August 8, 1995, from James M. Taylor, Executive Director for Operations,
NRC, to T.S. Kress, Chairman, ACRS, Subject: Response to ACRS Comments on
Commission Paper on Technical Issues Pertaining to the Westinghouse AP600 Design
. August 14, 1996
The Honorable Shirley Ann Jackson
Chairman
U.S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT: DESIGN CHANGES PROPOSED BY ASEA BROWN BOVERI - COMBUSTION
ENGINEERING RELATING TO THE CERTIFICATION OF THE SYSTEM 80+
DESIGN
During the 433rd meeting of the Advisory Committee on Reactor Safeguards, August 8-10, 1996,
we reviewed recent design changes proposed by ASEA Brown Boveri - Combustion Engineering
(ABB-CE) relating to the certification of the System 80+ design. These "design changes" consist of
both actual modifications to the design and corrections to the documentation to remove
inconsistencies and typographical errors. We had the benefit of discussions with representatives of
the NRC staff and of ABB-CE. We also had the benefit of the documents referenced.
Conclusions
Our review of Supplement 1 to NUREG-1462, "Final Safety Evaluation Report Related to the
Certification of the System 80+ Design," did not change the conclusion reached in our earlier report
of May 11, 1994. We continue to believe that acceptable bases and requirements have been
established in the application to assure that the System 80+ Standard Design can be used to engineer
and construct plants that with reasonable assurance can be operated without undue risk to the health
and safety of the public.
Background and Discussion
We have been involved in the review of the System 80+ design since ABB-CE applied for
certification. This review was carried out in accordance with 10 CFR Part 52, which requires ACRS
to report on those portions of 10 CFR Part 52 applications that concern safety. In our May 11, 1994
report to the Commission, we supported the certification of the System 80+ design. This report was
included in the staff Safety Evaluation Report (NUREG-1462). The present review is intended to
supplement our earlier review of this ABB-CE application.
Sincerely,
/s/
T. S. Kress
Chairman, ACRS
References:
1. U. S. Nuclear Regulatory Commission, NUREG-1462, Supplement No. 1, "Final Safety
Evaluation Report Related to the Certification of the System 80+ Design," dated July 1, 1996
2. ACRS Report dated May 11, 1994, from T. S. Kress, Chairman, ACRS, to Ivan Selin,
Chairman, NRC, Subject: Report on the Safety Aspects of the ASEA Brown Boveri-Combustion Engineering Application for Certification of the System 80+ Standard Plant
Design
3. Letter dated June 27, 1996, from C. B. Brinkman, ABB-Combustion Engineering Nuclear
Systems, to U.S. Nuclear Regulatory Commission, regarding System 80+ Standard Plant
Design Changes
4. Letter dated July 17, 1996, from C. B. Brinkman, ABB-Combustion Engineering Nuclear
Systems, to U.S. Nuclear Regulatory Commission, regarding six additional design changes
for System 80+ Standard Plant Design
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