United States Nuclear Regulatory Commission

Office of Public Affairs

Washington, DC 20555

Phone 301-415-8200 Fax 301-415-2234

Internet:opa@nrc.gov

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, performancebased 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 plantspecific 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

ˇ 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 nonsafety 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 designbasis and severe accidents. We recommend Commission approval.

ˇ 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.

ˇ 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, ABBCombustion 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, ABBCombustion Engineering Nuclear Systems, to U.S. Nuclear

Regulatory Commission, regarding six additional design changes

for System 80+ Standard Plant Design