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. 96-63 FOR IMMEDIATE RELEASE

(Friday, April 26, 1996)

NOTE TO EDITORS:

The Nuclear Regulatory Commission has received four attached

reports from its Advisory Committee on Reactor Safeguards (ACRS).

The reports, in the form of letters, provide comments on:

--Proposed revisions to NRC regulations and regulatory

guides relating to reactor site criteria.

--A probabilistic risk assessment (PRA) document, its pilot

applications, and the next steps to expand the use of PRA in the

regulatory decision-making process.

--Westinghouse Electric's best-estimate computer code for a

large-break loss-of-coolant accident.

--Continued need for U.S. membership in the Nuclear Energy

Agency, a part of the Organization for Economic Cooperation and

Development.

#

Attachments:

As stated
April 22, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: PROPOSED REVISIONS TO 10 CFR PARTS 50 AND 100 AND

PROPOSED REGULATORY GUIDES RELATING TO REACTOR SITE

CRITERIA

During the 430th meeting of the Advisory Committee on Reactor

Safeguards, April 11-13, 1996, we reviewed the proposed revisions

to reactor siting regulations and associated Regulatory Guides

and Standard Review Plan sections. Our Subcommittee on Extreme

External Phenomena reviewed this matter during a meeting on April

3, 1996. During this review, we had the benefit of discussions

with representatives of the NRC staff, Westinghouse Electric

Corporation, and the Nuclear Energy Institute. We also had the

benefit of the document referenced.

The staff has proposed final revisions to 10 CFR Parts 50 and 100

and a new Appendix S to Part 50 that deal with both seismic and

source term issues for future plants and sites. Many of the

implementation details will be found in new Regulatory Guides and

in Standard Review Plan sections. The existing requirements of

10 CFR Part 100 and its Appendix A will remain in effect for

operating plants.

We recommend that the proposed final rule dealing with the

seismic aspects be issued.

The proposed final rule requires that any individual, located at

any point on the exclusion area boundary for any two-hour period

following the postulated release of the fission products, not

receive a radiation dose in excess of 25 rem total effective dose

equivalent (TEDE). Similarly, an individual located at the outer

boundary of the low population zone (LPZ), who is exposed to the

radioactive cloud resulting from the release of the postulated

fission products (during the entire period of its passage), not

receive a dose in excess of 25 rem TEDE. Consistency within the

body of NRC regulations is most desirable. We recommend that

careful definitions of the TEDE limits that are mindful of organ

dose weighting factors found in 10 CFR Part 20 be included in the

final rule.
Radiological doses are to be evaluated over a two-hour period.

The proposed final rule states that the evaluation should be over

the two-hour period of maximum dose. The Office of Nuclear

Regulatory Research (RES) has a differing view and recommends

that the proposed final rule be modified from any two-hour period

after release of fission products (referred to as the "worst" two

hours) to a period of two hours commencing with fuel failure

(referred to as the "first" two hours). RES believes that the

use of the worst two-hour period in the dose calculation is not

justified by risk considerations and could lead to increased

costs for future licensees with no commensurate gain in safety.

The staff supporting the proposed rule states that (1) the

proposed licensing framework would provide a relaxation of

engineered safety feature (ESF) performance requirements

commensurate with updated source term and radiological insights,

(2) the regulatory requirements for determination of incontainment radioactive material during the two-hour dose

evaluation period would be consistent and capable of handling

designs substantially different from those analyzed in NUREG1465, "Accident Source Terms for Light-Water Nuclear Power

Plants," (3) the analysis would be easy to perform and

reproducible with confidence, and (4) the technical bases and

analytical methods would be defensible. While the revised dose

evaluation in 10 CFR 50.34 is intended for future plants, the

staff is concerned that a current licensee might seek to use it

to remove or disable existing fission product cleanup systems.

This could markedly change the risk profile of the plant from

that which was licensed.

We are not persuaded by the rationale provided by RES in favor of

the first two-hour dose calculation. We agree with the position

taken in the proposed final rule, and recommend that the rule and

the associated Regulatory Guides and SRP sections be issued.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

REFERENCE:

Memorandum dated March 6, 1996, from T. P. Speis, Office of

Nuclear Regulatory Research, NRC, to J. T. Larkins, ACRS,

transmitting Revisions to 10 CFR Part 100, Reactor Site

Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50

(Final Rules) and Associated Regulatory Guides and Standard

Review Plan Sections
April 23, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: PROBABILISTIC RISK ASSESSMENT FRAMEWORK, PILOT

APPLICATIONS, AND NEXT STEPS TO EXPAND THE USE

OF PRA IN THE REGULATORY DECISION-MAKING PROCESS

During the 430th meeting of the Advisory Committee on Reactor

Safeguards, April 11-13, 1996, we continued our deliberations on

risk-informed and performance-oriented regulation (RIPOR). We

met with representatives of the NRC staff and the Nuclear Energy

Institute (NEI) during our 429th meeting on March 7-9, 1996. Our

Subcommittee on Probabilistic Risk Assessment (PRA) also met on

October 26-27, 1995, with representatives of the NRC staff and of

the nuclear industry, and on February 27-28, 1996, with the NRC

staff and two invited experts, Dr. D. M. Karydas (performancebased standards for fire protection) and Professor T. G.

Theofanous (on the proper formulation of safety goals and

assessment of safety margins for rare and high-consequence

hazards). We also had the benefit of the documents referenced.

This report is in response to the Staff Requirements Memorandum

dated December 27, 1995, in which the Commission requested "ACRS

views on the PRA framework document, its relationship to the

pilot applications (SECY-95-280), and the next steps in the

process to expand the use of PRA in the regulatory

decision-making process."

PRA Framework Document

The PRA framework document provides a good starting point in the

development of RIPOR. The six-step process described in the

document is a reasonable way to proceed. We agree with the staff

that the focus should be on the integration of probabilistic and

deterministic approaches to regulation.

The PRA framework document, however, does not articulate an

overall philosophy for RIPOR. We believe that such a philosophy

should be developed. Some important high-level principles that

should be included are:

1. RIPOR should consider risk from all modes of nuclear plant

operations, including full power, shutdown, and transition.

2. The Commission's safety goals should serve as the top-level

acceptance criteria.
3. Subsidiary performance-based acceptance criteria should be

determined in a consistent way and must be measurable or

calculable. The licensee should be granted flexibility in

choosing the means to meet the criteria.

4. The relationship between RIPOR and defense-in-depth should

be explained. The role of defense-in-depth in the determination of performance criteria to accommodate uncertainty and

incompleteness in risk assessments should be established.

5. Criteria for the adoption of prescriptive regulations should

be clearly delineated.

6. The acceptance criteria should be set at the highest level

of plant system hierarchy that is consistent with the other

principles noted above.

Discussion

It is indicative of the novelty of these concepts that we have

spent a considerable amount of time discussing the meaning of

"performance" among ourselves and with the staff and NEI. Some

interpret performance in a limited way; i.e., its measures are

simply the reliability and availability (or related quantities)

of plant systems and components. Others take a broader view and

interpret it as the overall performance of the licensee,

including operations, maintenance, training, and the prevailing

safety culture at the plant.

Similarly, the definition of performance criteria varies widely.

At one extreme, we have simple measures that are either directly

measurable or that involve calculations (e.g., the reliabilities

and unavailabilities mentioned above). At the other extreme,

performance criteria can be probabilistic or nonprobabilistic and

can be set at any level. Observations and statistical or

experimental evidence from the plant or other sources in

conjunction with models can be used to demonstrate that the

criteria have been met. As part of an overall philosophy, the

staff needs to resolve the ambiguity in the definition of

performance criteria.

Pilot Applications

While we support the staff's use of pilot applications, we are

concerned that there seems to be no integrated justification for

their selection. We would like to see the development of a list

of important issues that are expected to arise on the road to

RIPOR,
along with a discussion of how the selected pilot projects will

help. The staff has agreed to look into these issues.

We also recommend that, for each pilot project, attempts be made

to establish performance-based decision criteria along with the

methods that would be used for demonstrating compliance. Such an

exercise should provide useful insights regarding the overall

feasibility of a performance-oriented approach to regulation.

Next Steps to Expand the Use of PRA in the Regulatory Decisionmaking Process

We believe that the NRC needs to take a number of important

additional steps before a RIPOR environment can be achieved.

These are discussed below.

Safety Goals

A restatement of the Commission's safety goal policy is needed

that will allow the use of safety goals on a plant-specific

basis.

Performance-Based Regulatory Criteria

A methodology is needed to determine performance-based criteria

for regulatory action that are consistent with the top-level

safety goals, as stated in the high-level principles. A "topdown" approach will ensure that this happens. An important

element should be the preservation of the concept of defense-indepth. The development of this methodology will also provide the

opportunity to reexamine the validity of Level 2 subsidiary

goals, which appear to be controversial at this time.

Programmatic Issues

Developing a RIPOR system should be a participative effort

between the staff and the industry. We believe that the

magnitude and significance of the task that the staff has

undertaken requires a cooperative effort. Also, we recommend

that the staff work with foreign researchers and regulatory

agencies.

Conclusion

The intellectual and practical issues that the staff must

confront in developing a RIPOR structure are significant. The

staff has made a good start, but much remains to be done. We are

pleased that the staff has agreed to meet with us periodically.

Recent meetings have demonstrated that the staff is receptive to

suggestions on how to deal with these complex issues. We applaud

this attitude. We will keep you informed as these efforts

progress.
Additional comments by ACRS Members Thomas S. Kress and Don W.

Miller are presented below.

Sincerely,

/S/

T. S. Kress

Chairman, ACRS

Additional Comments by ACRS Members Thomas S. Kress and Don W.

Miller

While we agree with most of the Committee's report on this

subject, we find it to lack coherence. The major problem we have

with the Committee report is its treatment of the concept of

"performance-based" regulation. We conceive of basically two

meanings to the word "performance" in this context: (1) the

performance of equipment (systems and components) in carrying out

the intended function, or (2) the performance of the licensee in

performing its function (operation, maintenance, inspection,

training, etc.). The first of these could further relate to

either the operability of the specific equipment (e.g., does it

turn on or off, and, in the case of a pump, for example, does it

provide the required flow) or to the reliability/availability of

the equipment. In our view, the former does not provide any

basis on which to develop a regulatory structure (there are no

meaningful acceptance criteria that relate to risk). On the

other hand, the latter can clearly be anchored in risk. This,

however, would be purely risk-based regulation. The word

"performance" in this context becomes synonymous with "risk" and

such a regulatory concept should be designated as risk-based and

should not be called performance-based.

The second possible meaning of performance, the performance of

the licensee, obviously has a nexus to risk. This connotation of

performance, however, is what we have been calling organizational

factors. To date, a methodology has not been developed by which

objective performance measures can be identified and be factored

directly into PRA to quantify risk implications. Therefore, at

this time, we do not have the capability to develop such

performance-based regulations in any coherent manner. This

would, however, be an area worth pursuing in the future with

additional research.

This leads us to our main point. At this time, we should be

striving for risk-based or risk-informed regulations and should

relegate the concept of "performance" regulation to being a

remote possibility that needs substantial research to determine

feasibility.
References:

1. Memorandum dated December 27, 1995, from J. Hoyle, Secretary

of NRC, to J. Larkins, ACRS, Subject: Staff Requirements

Memorandum dated December 27, 1995

2. Memorandum dated June 16, 1995, from A. Bates, Office of the

Secretary, NRC, to File, Subject: Staff Requirements

Memorandum dated June 16, 1995

3. Letter dated February 6, 1996, from J. Milhoan, Office of

the Executive Director for Operations, NRC, to W. Rasin,

Nuclear Energy Institute, Subject: Improving the Regulatory

Process through Risk-Based and Performance-Based Regulation

4. Letter dated January 3, 1996, from J. Taylor, Executive

Director for Operations, NRC, to Chairman Jackson, NRC,

Subject: Improvements Associated With Managing the

Utilization of Probabilistic Risk Assessment (PRA) and

Digital Instrumentation and Control Technology

5. Letter dated November 30, 1995, from Chairman Jackson, NRC,

to J. Taylor, Executive Director for Operations, NRC,

Subject: Follow-up Requests in Probabilistic Risk Assessment

and Digital Instrumentation and Control

6. SECY-95-280, "Framework for Applying Probabilistic Risk

Analysis in Reactor Regulation," dated November 27, 1995

7. Letter dated November 14, 1995, from W. Rasin, Nuclear

Energy Institute, to J. Milhoan, Office of Executive

Director for Operations, NRC, Subject: Draft report,

"Improving the Regulatory Process Through Risk-Based and

Performance-Based Regulation"
April 19, 1996

The Honorable Shirley Ann Jackson

Chairman

U.S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: WESTINGHOUSE BEST-ESTIMATE LOSS-OF-COOLANT ACCIDENT

ANALYSIS METHODOLOGY

During the 430th meeting of the Advisory Committee on Reactor

Safeguards, April 11-13, 1996, we concluded our review of the

best-estimate, large-break, loss-of-coolant accident (LBLOCA)

analysis methodology developed by the Westinghouse Electric

Corporation. We had previously reviewed this matter during our

428th meeting, February 8-10, 1996. We also had the benefit of

the referenced documents.

In our February 23, 1996 report commenting on the results of our

initial review, we identified several technical details of the

Westinghouse LBLOCA methodology needing further attention and

also commented on the adequacy of the documentation. As a result

of subsequent discussions with representatives of Westinghouse

and the NRC staff during this meeting, we believe that these

concerns have been addressed.

ACRS Member George Apostolakis did not participate in the

Committee's deliberation of this matter.

Sincerely,

/s/

T. S. Kress

Chairman, ACRS

References:

1. Memorandum dated March 25, 1996, from N. Liparulo,

Westinghouse, to Nuclear Regulatory Commission, transmitting

information on the resolution of issues related to the

review of WCAP-12945-P (Proprietary)

2. Letter dated March 15, 1996, from J. Taylor, Executive

Director for Operations, NRC, to T. S. Kress, Chairman,

ACRS, Subject: Westinghouse Best-Estimate Loss-of-Coolant

Accident Analysis Methodology

3. Report dated February 23, 1996, from T. S. Kress, Chairman,

ACRS, to Shirley Ann Jackson, Chairman, NRC, Subject:

Westinghouse Best-Estimate Loss-of-Coolant Accident Analysis

Methodology
April 17, 1996

The Honorable Shirley Ann Jackson

Chairman

U. S. Nuclear Regulatory Commission

Washington, D.C. 20555-0001

Dear Chairman Jackson:

SUBJECT: CONTINUED NEED FOR UNITED STATES MEMBERSHIP IN THE

NUCLEAR ENERGY AGENCY

The Advisory Committee on Reactor Safeguards has recently learned

of the proposed withdrawal of the United States (U.S.) from

participation in the Nuclear Energy Agency (NEA), a part of the

Organization for Economic Cooperation and Development. The

Department of Energy (DOE) and the Nuclear Regulatory Commission

(NRC) are the primary U.S. technical participants in the NEA

activities and, hence, are the agencies that have the most

complete understanding of the benefits of membership in NEA. Our

comments will perhaps assist you as you set forth the NRC

position.

For many years, the NEA Committee on Safety of Nuclear

Installations (CSNI) has been an active and productive leader in

nuclear reactor safety research. CSNI reports cover the full

scope of reactor safety concerns and are prepared by leading

technical experts from the primary technical research

laboratories and agencies of the member countries.

Current CSNI efforts contribute to U.S. programs in extended

burnup reactor fuels, high-pressure melt ejection, direct

containment heating, accident management, and steam explosions.

Clearly, the CSNI has demonstrated the ability to keep pace with

real concerns in nuclear safety. Furthermore, these efforts have

resulted in substantial savings in U.S. research costs.

Nuclear safety is truly an international concern. The NEA is a

forum for the consideration of common technical safety issues by

the responsible regulatory agencies in the member countries and

has been useful in developing consistent "western" positions. If

the NEA did not exist, we would soon be convinced that it should

be invented.
We believe that the suggested U.S. withdrawal from the NEA is

shortsighted. We fully support your efforts to ensure continued

U.S. participation in the NEA.

Sincerely,

/S/

T. S. Kress

Chairman, ACRS