ML20151V546

A Compilation of Reports of the Advisory Committee on Reactor SAFEGUARDS.1987 Annual

ML20151V546
Person / Time
Issue date:	04/30/1988
From:	Advisory Committee on Reactor Safeguards
To:
References
NUREG-1125, NUREG-1125-V09, NUREG-1125-V9, NUDOCS 8805030096
Download: ML20151V546 (137)

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l NUREG-1125

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Volume 9 (k, /

A Compiation of Reports of The Advisory Committee on Reactor l

Safeguards i

1987 Annual U.S. Nuclear Regulatory Commission l

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April 1988 i

j G805030096 880430 PDR NUREG 1185 R PDR i

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NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in NRC publications will be available from one of the following sources:

1. The NRC Public Document Room,1717 H Street, N.W.

Washington, DC 20555

2. The Superintendent of Documents, U.S. Governmeat Printing Office, Post Office Box 37082, Washington, DC 20013 7082

3. The National Technical Information Service, Springfield, VA 22161 Although the listing that follows represents the majority of documents cited in NRC publications, j

lt is not intended to be exhaustive.

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Refersnced documents available for inspection and copying for a fee from the NRC Public Docu-

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ment Room include NRC correspondence and internal NRC memoranda; NRC Office of Inspection and Enforcement bulletins, circulars, information notices, inspection and investigation notices; Licensee Event Reports; vendor reports and correspondence; Commission papers; and applicant and licensee documents and correspondence.

The following documents in the NUREG series are available for purchase from th2 GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proc)edings, and NRC booklets and broghures. Also available are Regulatory Guides, NRC regulations in the Code of i

Federal Regulations, and Nuclear Regulatory Commission issuances.

Documents available from the National Technical Information Service include NUREG series reports and technical reports prepared by other federal agencies ar.d reports prepared by the Atomic 1

Energy Commission, forerunner agency to the Nuclear Regulatory Commission.

Documents available from public and special technical libraries include all open literature items, such as books, joumal and periodical articles, and transactions. Federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries, i

Documents such as theses, dissertations, foreign reports and translations, and non NRC conference proceedings are available for purchase from the organization sponsoring the publication cited.

Single copies of NRC draf t reports are available free, to the extent of supply, upon written request to the Division of Information Support Services, Distribution Section, U.S. Nuclear Regulatory Commission, Washington, DC 20555.

l Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library,7920 Norfolk Avenue, Bethesda, Maryland, and are available there for reference use by the public. Codes and standards are usually copyrighted and may be purchased from the originating organization or, if they are American National Stt,ndards, from the American National Standards Institute,1430 Broadway, New York, NY 10018.

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ABSTRACT This compilation contains 47 ACRS reports submitted to the Comission or to the Executive Director for Operations during calendar year 1987.

It also includes a report to the Congress on the NRC Safety Research Program for FY 1988 All reports have been made available to the public through the NRC Public Document Room and the U.S.

Library of Congress.

The reports are divided into two groups:

Part 1: ACRS Reports on Project Reviews, and Part 2:

ACRS Reports on Generic Subjects.

Part I contains ACRS reports alphabetized by project name and within project name by chronological order.

Part 2 categorizes the reports by the most appropriate generic subject area and within subject area by chronological order.

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_ PREFACE The enclosed reports represent the recernendations and comments of the U.S.

Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards during calendar year 1987.

This publication Volume 9, is an annual supplenent to NUREG-1125.

Previous issues of NUREG-1125 are as follows:

Volume Inclusive Dates 1 through 6 September 1957 through December 1984 7

Calendar Yea

• 1985 8

Calendar Year 1986 l

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ACRS MEMBERSHIP (1987)

CHAIRMAN:

Dr. William Kerr University of Michigan VICE C'4 AIRMAN:

Dr. Forrest J. Remick The Pennsylvania State University MEf!PERS:

Dr. Max W. Carbon University of Wisconsin Mr. Jesse C. Ebersole, Petired Tennessee Valley Authority Dr. Harold W. Lewis University of California, Santa Barbara Dr. J. Carson Mark, Retired Los Alamos Scientific Laboratory Mr. Carlyle Michelson, Petired Tennessee Valley Authority and AE00/US Nuclear Regulatory Comission Dr. Dade W. Moeller Harvard University Dr. David Okrent University of California, Los Angeles Mr. Glenn A. Reed, Retired Point Beach Nuclear Power Plant Wisconsin Electric Power Company Dr. Paul G. Shewmon Ohio State University Dr. Chester P. Siess, Retired (Prof. Emeritus)

University oT Illinois Dr. Martin J. Steindler Argonne National Laboratory Mr. David A. Ward E.I. du Pont de Nemours & Company Savannah River Laboratory Hr. Charles J. Wylie, Retired Duke Power Company vii a

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f TABLE OF CONTENTS VOLUME IX i

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ABSTRACT..............................................................

iii PREFACE...............................................................

Y MEMBERSHIP............................................................ Vii PART 1: ACRS REPORTS ON PROJECT REVIEWS Seabrook Station, Units 1 and 2.....................................

1 Weqtinghouse RESAR/SP-90 Design.....................................

5 PART 2: ACRS REPORTS ON GENERIC SUBJECTS Advanced Reactors ACRS Coments on Draf t NUREG-1226. "Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants "

June 9, 1987................................,......................

7 Ay iliary Systems ACRS Coments on Nuclear Power Plant Air Cooling Systems. October 15, 1987........................................

11 Class 9 Accidents ACRS Coments on the Draft NRC Peport on the Implications of the Accident at Chernobyl Nuclear Station Unit 4. January 14, 1987.........................

13 ACR$ Coments on the Implications of the Accident at the Chernobyl Nuclear Station Unit 4. January 15, 1987.........................................

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TABLEOFCONTENTS(CONT'D)

Page Class 9 Accidents (cont'd)

ACRS Report on Proposed Research to Reduce Source Term Uncertainty, May 13, 1987............................

17 ACRS Report on Proposed Generic Letter on Individual Plant Examinations for Severe Accident Vulnerabilities, June 9, 1987...........................

21 Emergency Core Cooling Systems ACRS Comments on Code Scaling, Applicability and Uncertainty Methodology for Determination of Uncertainty Associated with the Use of Realistic ECCS Evaluation Models, September 16, 1987........................

27 Emergency Plar.ning ACRS Comments on Developments in Emergency Planning, September 16, 1987.....................................

31 Engineered Safeguards ACRS Comments on Testing of Charcoal Adsorption Capacity, February 10, 1987......................................

33 ACRS Comments on Research into Continuous Leakage Monitoring, July 16, 1987................................

35 Fire Protection Preliminary ACRS Views en Fire Risk Research Scoping Study August 10, 1987...................................

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i TABLEOFCONTENTS(CONT'D)

Page Generic Issues / Unresolved Safety Issues ACRS Comment on Prioritization of Generic Issue 61:

"SRV Discharge Line Break Inside the Wetwell Air-Space of BWR Mark I and Mark II Containments,"

March 10, 1987...................................................

39 ACRS Comments on the NRC Staff Proposal for the Resolution of USI A-44, "Station Blackout,"

June 9, 1987.....................................................

41 ACRS Comments on the Proposed Resolution for Generic Issue 124, "Auxiliary Feedwater System Reliability,"

October 15, 1987.................................................

45 Human Factors ACRS Comments on Licensee Event Reports Pertaining to Control Room Habitability, July 15, 1987......................

49 Metal Components ACRS Comments on the Embrittlement of Structural Steel, July 15, 1987,............................................

51 ACRS Comments on Memorandum from Victor Stello, Jr., EDO, Dated October 7,1987 Regarding the Embrittlement of Structural Steel, December 8, 1987..............

53 Operating Organizations l

ACRS Coerents on the Advance Notice of Proposed Rulemaking:

Degree Requirements for Senior

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Operators, August 12, 1987.......................................

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1 TABLE OF CONTENTS (CONT'D)

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Quality Assurance / Quality Control ProposedinternationalWorkshoponQualityin Design and Construction of Nuclear Power Plants, i

June 10, 1987....................................................

59 ACRS Coments Regarding Proposed International Werkshop on Quality in Design and Construction of Nuclear Power Plants, September 15, 1907.........................

61 Radiological Effects ACRS Coments on International Cooperation en Research Related to Radiation Protection, July 15, 1987....................................................

65 ACPS Coments on the Development of Radiation Protection Standar/

November 10, 1987..........................

67 Reactor Safety Development ACRS Recommendations on Improved Safety for Future Light Water Reactor Plant Design, January 15, 1987.................................................

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i ACPS Coments on Improved Safety for Future

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Light Water Reactors, July 15, 1987.....................

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77 ACRS Coments on Improved Safety for Future l

Light Water Reactors, September 15, 1987.........................

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TABLE OF CONTENTS (CONT'D)

Page Regulatory Guides ACRS Action on the Proposed Regulatory Guide EE 404-4, "Environmental Qualification of' Connection Assemblies for Nuclear Power Plants,"

April 13, 1987 '...................................................

81 ACRS Action on Proposed Revision 2 of Regulatory Guide 1.100, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants,"

June 9, 1987.....................................................

83 ACRS Action on Proposed Regulatory Guide EE 006-5, "Qualification of Safety-Related Lead Storage Batteries for Nuclear Powe~r Plants," August 11, 1907.............

85 ACRS Action on Proposed Final Regulatory Guide (Task EE 404-4), "Environmental Qualification of Connection Assemblies for Nuclear Power Plants,"

October 14, 1987.................................................

87 Reliability and Probabilistic Analysis ACRS Comments on an Implementation Plan for the Safety Goal Policy, May 13, 1987.................................

89 ACRS Comments on Draft NUREG-1150, "Reactor Risk Reference Document," July 15, 1987...............................

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TABLEOFCONTENTS(CONT'D)

Page Rules and Regulations ACRS Coments on Proposed Revised Standard Review Plan Section 3.6.2, "Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Ruptut e of Piping," dated October 2,1986, -

June 9, 1987.....................................................

103 ACRS Coments on Proposed Revisions to Standard Review Plan Sections 6.5.2, "Containment Spray as a Fission Product Cleanup System" and 6.5.5, "Suppression Pools as Fission Product Cleanup Systems," June 9, 1987...........................................

105 ACRS Action on the Proposed Section 3.6.3, "Leak-Before-Break Evaluation Procedures," of the NRR Standard Review Plan, August 12, 1987....................

107 ACRS Coments on Proposed Final Broad Scope Rule to Modify General Design Criterion 4, Environmental and Missile Design Bases (GDC-4), August 12, 1987................

109 ACRS Coments on the Revised ASB 31 of Standard Review Plan 3.6.1, Requirements for Arbitrarily Postulated Jet Impingement Effects in the Break Exclusion Zone, December 8, 1987.................................

111 Safety Research Review and Evaluation of the Nuclear Regulatory Comission's Safety Research Program for Fiscal Year 1988, February 11, 1987.....................................

113 ACRS Reports to the NRC on the Research Program, April 13, 1987...................................................

123 Systematic Evaluation ACRS Coments on the Integrated Safety Assessment Program, July 15, 1987...........................................

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TABLE OF CONTENTS (CONT'D)

Page Waste Management

- ACRS Comments on the Advance Notice of Proposed Rulemaking on the Definition of "High-Level Radioactive Waste," March 9, 1987................................

135 ACRS Comments on "Standard Format and Content" (NUREG-1199) and "Standard Review Plan" (NUREG-1200),

Guidance Documents for the Preparation of a License Application for a Low-Level Waste Disposal Facility, March 9, 1987....................................................

137 ACRS Comments on Proposed Nuclear Waste Advisory Committee, April 14, 1987........................................

139 ACRS Comments on Disposal of Mixed Waste, June 9, 1987.....................................................

141 ACRS Comments on Quality Assurance Programs for a High-Level Waste Repository, June 10, 1987.....................

143 ACRS Action on the Proposed Final Rule Amendments to 10 CFR Part 72, "Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste," September 17, 1987.....................

145 ACRS Comments on Radioactive Waste Management Research and Other Activities, November 10, 1987.................

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NUCLEAR REGULATORY COMMISSION g

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WASHINGTON, D. C. 20555 e....

March 9, 1987 The Honorable Edward J. iiarkey Comittee on Eneray and Comerce United States House of Representatives Washington, D.C.

20515

Dear Ccngressrian Markey:

We note your interest in our ongoing deliberations relative to the Seabrook Station, as evidenced by your letter of February 26, 1987 to Mr. David A.

Ward, ACRS.

Section 182b of the Atomic Energy Act requires the Advisory Comittee on Reactor Safeauards to "review each application for... an operatino license for a facility...." The Comittee issued a report, dated April 19, 1983, with respect to the proposal to operate the Seabrook Station; a copy of that report is attached.

In the report, we indicated that there were some open issues, and we noted the absence of a fully developed emergency plan.

Because the Comittee reported a satisfactory conclusion only with respect to operation of the plant at power levels at or below 5 percent of full power, our review of the Seabrook Station operating license is not compl e te.

We have begun a review of matters associated with emergency planning for the Seabrook Station.

When we have completed our work, and fulfilled our obligation to provide sound and dispassionate advice to the Comission, that advice will be publicly available, as will the listing of inputs that contributed to it.

We will provide you with a copy of our report at that tir.e.

1 Sincerely, William Kerr Chairman

Attachment:

Letter from J.C. Ebersole, Acting Chairman, ACRS, to N.J. Palladino, Chairman, NRC, dated April 19, 1983 cc: Honorable Philip R. Sharp, Chairman Subcomittee on Energy and Power 1

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,o'g April 19, 1983 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, DC 20555 l

SUBJECT:

ACRS REPORT ON LOW POWER OPERATION OF THE SEABROOK STATION, UNITS 1 AND 2 During its 276th meeting, April 14-16, 1983, the Advisory Committee on Reactor Safeguards reviewed the application of the Public Servi.ce Company of New Hampshire, acting as agent for and on behalf of the Seabrook Owners Group (the Applicant), for an operating license for the Seabrook Station, Units 1 and 2.

The station is to be operated by the Public Service Company of New Hampshire.

This application was considered at an ACRS Subcommittee meeting in Hampton Beach, New Hampshire, on April 1-2, 1983. Members of the Subcommittee toured the facility on April 1,1983.

In our review, we had the benefit of discussions with representatives of the Applicant, the Yankee Atomic Electric Company, Westinghouse Electric Corporation, United Engineers and Constructors, Inc., the NRC Staff, and with members of the publ i c. We also had the benefit of the documents listed below. The Commit-tee commented on the construction permit application for Seabrook Station, Units 1 and 2 in a report dated December 10, 1974.

The Seabrook Station is located on the western side of Hampton Harbor, in the Township of Seabrook, Rockingham County, New Hampshire, approximately 11 miles south of Portsmouth, New Hampshire and 40 miles north of Boston, Massachusetts.

Each Seabrook unit uses a Westinghouse nuclear steam supply system with a rated core power of 3411 MWt.

The containment for each unit consists of a steel lined, reinforced concrete structure which is surrcunded by-a reinforced concrete containment enclosure.

The design pressure of the containment is 52 psig.

The annular space between containment and enclosure is maintained at a slight negative pressure.

Seabrook will use Westinghouse Model F steam generators, which incorporate design changes intended to eliminate the problems experienced with earlier model s.

We wish to be kept informed concerning the performance of these steam generators.

We were favorably impressed by the amount of attention given and resources expended in the area of personnel training.

The result appears to be an i

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Honorable Nunzio J. Palladino April 19,1983 excellent educational system for operations personnel, including operators and technicians.

The resources at the disposal of the Applicant, including those of the Yankee Atomic Electric Company, appear to be appropriate for the operation of this nuclear power sta_ tion.

The ACRS has on several occasions recommended that evaluations be made of the capability of light water nuclear power plants to be shut down safely in the event of an earthquake of greater severity and lower likelihood than the safe shutdown earthquake.

The implications of recent seismic activity, such as the January 1982 earthquakes in central New Brunswick and New Hampshire, are being evaluated.

We recommend for the Seabrook Station that specific attention be given to the seismic capability of those components that are important to the accomplishment of safe shutdown including the emergency AC power supplies, the DC power supplies, and small components such as actuators and instrument lines.

The Applicant has undertaken a full-scope probabilistic risk assessment (PRA) which is scheduled for completion about October 1983.

The ACRS wishes to be kept informed concerning the results of the NRC Staff's review and evaluation of this PRA.

L The Seabrook Station, Units 1 and ? will be the first commercial nuclear power plant in the state of New Hampshire; the Station is also situated very close to the New Hampshire-Massachusetts border.

As a result, the NRC Staff and Applicant must give particular attention to assuring proper coordination with appropriate state and regional agencies in the development of effective emergency plans.

There is a large summertime increase in population within a few miles of the site due to the beach areas of Seabrook and Hampton, New Hampshire.

The nature of the road network serving the beach requires that special attention be given to the problems associated with evacuation.

Because the emergency plan is not yet fully developed, we were unable to review it.

A number of other items have been identified by the NRC Staff as Outstanding Issues. There is also a set of Confinnatory Issues that awaits additional documentation.

We found no reason to believe that any of these issues will be especially difficult to resolve. We recommend that they be resolved in a l

manner satisfactory to the NRC Staff.

l Fuel loading for Unit 1 is scheduled for September 1984 and fuel loading for Unit 2 is planned to take place about 2.5 years after fuel loading for Unit 1.

Should there be a significant delay in this schedule, we would expect to examine the need for additional review of Unit 2.

We believe that, if due regard is given to the items mentioned above, and subject to satisfactory completion of construction, staf fing, and preoperational testing, there is reasonable assurance that the Seabrook 3

E Honorable,Nunzio J. Palladino April 19,1983 Station, Units 1 and 2, can be operated - at core power levels up to 5 per-cent of full power without undue risk to the health and safety of the public.

Sincerely, 4

Jesse C. Ebersole Acting Chainnan

References:

1. Public Service Company of New Hampshire, Seabrook Station "Final Safety Analysis Report," Volumes 1-15, with Amendments 45-48

2. U.S. Nuclear Regulatory Commission, "Safety Evaluation Report Related to the Operation of Seabrook Station, Units 1 and 2," NUREG-0896, dated March 1983.

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3. Written Public Comments from J. Doughty, Seacoast Anti-Pollution League (SAPL),

Subject:

SAPL Comments to the Advisory Committee on Reactor Safeguards Subcommittee Conducting the Independent Technical Review for j

the Seabrook Nuclear Power Plant, April 1983, received April 1,1983.

4. Written Public Comments from Rep. Roberta C. Pevear, New Hampshire House.

of Representatives,

Subject:

Statement Before Advisory Committee on Reactor Safeguards Meeting on Seabrook Operating License, April 2,1983, received April 2, 1983.

5. Written Public Comments from Elizabeth Dolly Weinhold,

Subject:

Seismic Issues, received April 2, 1983.

6. Written Public Comments from Rep. Roberta C. Pevear, New Hampshire House of Representatives,

Subject:

Response to Kulash Report on evac :ation planning, dated April 4,1983.

7. Written Public Comments from Diana P. Sidebotham, President, New England Coalition on Nuclear Pollution, Inc.,

Subject:

Remarks Prepared for delivery at April 1,1983 Subcommittee meeting on Seabrook Station, dated April 11, 1983.

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UNITED STATES

' "f, NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o,,

a WASHINGTON, D. C. 20555 l

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August 11, 1987 I

The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear RpWlatory Commission Washington, D.C.

20555

SUBJECT:

ACRS REVIEW 0F APPLICATION FOR PRELIMINARY APPROVAL OF THE WESTIhtHOUSE RESAR/SP-90 DESIGN Ouring the 328th meeting of the ACRS, August 6-8, 1987, we heard a presentation by the NRC Staff describing its schedule for review of the subject application.

We regard this program as highly important and expect to participate exter.stvely in the review.

!!e believe it will be appropriate for the ACRS subcomittee on Westinghouse reactor plants to meet with the NRC Staff and the Westinghouse representatives in a series of meetings beginning well in advance of the date when a draft SER is expected to be available.

Among the subjects that we would like to review are the scope and results of the probabilistic risk assessment used by Westinghouse in support of its design; a comparison of the design with modern plants of similar type, both in the U.S. and abroad, for example, the Vogtle, Paluel, Sizewell B, and KONV0I plants and the Japanese APWR; and similar comparison with the EPRI requirements document for future PWRs to the extent feasible.

We will be interested in the reasons for design choices and the NRC Staff evaluations of these choices.

Sincerely, William Kerr Chairman S

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Part 2: ACRS Reports on Generic Subjects

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NUCLEAR REGULATORY COMMISSION n

I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 o% v...../

June 9, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Commission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON DRAFT NUREG-1226, "DEVELOPMENT AND UTILIZA-TION OF THE NRC POLICY STATEMENT ON THE REGULATION OF ADVANCED NUCLEAR POWER PLANTS" During the 326th neeting of the ACRS, June 4-6, 1987, and in our 325th meeting, May 7-9, 1987, we discussed NUREG-1226, "Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants." A Subconinittee macting was also held to discuss this NUREG with the NRC Staff on April 24, 1987.

During our discussion, we had the benefit of the documents referenced and also of earlier meetings with the NRC Staff.

We had previously reviewed the Advanced Reactor Policy Statement and had comented on the statement in a letter to Chairman Palladino dated October 16, 1985.

When the Advanced Reactor Policy Statement was issued, in July 1986, the Comission directed the NRC Staff to prepare a -document that would describe its development.

Later the purpose of the document (which became NUREG-1226) was extended to include factors important to implementation of the policy. Our comments will be limited to the implementation aspects of the document.

We are in general agreement with the implementation approach, but have several coments.

The early interactions between the Staff and an applicant are to be concerned with review of conceptual. design, well in advance of any formal application for a construction permit or a design certifica-tion.

The Staff reported that it intends to assure a conceptual design that looks ahead to possible future standardization.

We Concur.

The implementation plan encourages, but does not require, the devel-opment of new designs based on building and operation of prototypes.

We 'believe that operation of prototypes prior to certification of i

designs should be the norm and the only exceptions should be made in l

carefully evaluated cases, where there exists a sufficiently well-l developed experience base.

t The Honorable Lando W. 7.ech, Jr. June 9, 1987 NUREG-1226 uses the terms "defense-in-depth" and "design-basis accident."

These are time-honored terms, but they are inexact as concepts.

For example, there is a requirement to consider "beyond design basis" scenarios in the design. This presents, at minimum, a serious semantic problem.

We believe the Staff needs to clarify its use of these terms.

The policy statement encourages use of "performance-based" rather than "prescriptive" requirements.

Again we have concerns that these terms are used without being well defined.

For example, 10 CFR 50.46 is certainly a performance-based requirement for the design of an Emergency Core Cooling System (ECCS), but prescriptions for analyzing performance are given in excruciating detail in Appendix K.

We believe there is a need to clarify both of these terms and concepts.

We believe the attribute "simplicity" is not always a virtue to be encouraged in future nuclear power plants.

From the perspective of safety it is important to have plant systems designed to be easy to operate, easy to maintain, easy to understand, and capable of accom-modating a broad spectrum of challenges.

However, simplicity does not always provide these characteristics.

As an example, increased automation, as a means to make a plant easier to operate, may ac-tually make the design more compler.

The history of the evolution of engineered systems indicates they often become more complex as they are improved in reliability and performance, including' safety perfor-mance.

. We believe that NUREG-1226 should provide more definitive. guidance for sabotage-protection considerations for advanced plant designs.

We recognize this as a difficult issue, and it is for this reason that the Staff should give it additional attention.

Additional remarks by ACRS Member David Okrent are presented below.

Sincerely, William Kerr Chairman Additional Remarks by ACRS Member David Okrent I believe that defense-in-depth should be maintained such that an appropriate containment or other system intended to mitigate severe core melt accidents will be provided.

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The Honorable Lando W. Zech, Jr. June 9, 1987

References:

1.

U.S. Nuclear Regulatory Comission, NUREG-1226, "Development and Utilization of the NRC Policy Statement on the Regulation of Advanced Nuclear Power Plants," draft published May 5,1987.

2.

U.S.

Nuclear Regulatory Comission, SECY-85-279,

Subject:

"Revised Advanced Reactor Policy Statement," dated August 21, 1985.

3.

U.S.

fluclear Regulatory Comission, "Regulation of Advanced Nuclear Power Plants, Statement of Policy," 51 FR 24643, dated July 8, 1986.

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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0,

g WASHING TON, D. C. 20555 October 15, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON NUCLEAR POWER PLANT AIR COOLING SYSTEMS During the 330th meeting of the ACRS, October 8-10, 1987, we discussed a report from our Subcommittee on Auxiliary Systems regarding heating, venti-lating, and air conditioning system failures and their imoset on safety systems.

This matter was discussed on June 27, 1986 during o,ioint meeting of the ACRS Subcommittees on Occupational and Environmental Protection Systems and on Auxiliary Systems.

It was also discussed by the Auxiliary Systems Subcomittee during a meeting held on October 1,1987.

The Subcom-mittees had the benefit of discussions with representatives of the Office of Nuclear Reactor Regulation and the document referenced.

During the June 27, 1986 meeting, representatives of ths NRC Staff stated that failures of air cooling systems for areas houstag key components (for example, RHR pumps, switch gear, diesel generators, etc.) in certain nuclear power plants contribute significantly to estimated core-melt frequencies.

Because corrective measures are often taken once potential cooling system failures are identified, the impact of these potential failures on the proper functioning of these systems has not been reflected in the final PRAs issued for these plants.

As a result, sonie members of the NRC Staff and some licensees whose plants have similar deficiencies may not be aware of these problems.

Based on these observations, we recommend that the NRC Staff examine the extent to which these problems may be generic and take any corrective actions deemed necessary.

Sincerely.

William Kerr i

Chairman i

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Mr. Victor Stello, Jr. October 15, 1987

Reference:

Pris~entation material provided by Arthur Buslik, Office of Nuclear Reactor Regulation, before a joint meeting on June 27, 1986 of the ACRS Subcomittees on Occupational and Environmental Protection Systems and on Auxiliary Sys-tems.

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o UNITED STATES g

NUCLEAR REGULATORY COMMISSION o

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I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

WASHINGTON, D. C. 20555 J a n uary 14, 1987 l

Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555 l

SUBJECT:

ACRS COMMENTS ON THE DRAFT NRC REPORT ON THE IMPLICATIONS OF THE ACCIDENT AT CHERN0BYL NUCLEAR STATION UNIT 4 During the 321st meeting of the ACRS, January 8-10, 1987, we considered the implications of the accident at the Chernobyl nuclear station Unit 4 as it relates to nuclear power plants in the United States.

This subject was also considered during our 320th meeting, December 11-13, 1986 and our 319th meeting, Noverc.ber 6-8, 1986.. In our review, we had the benefit of meetings of our Subcomittee on Safety Philosophy, Technology, and Criteria held on November 5 and December 10, 1986, of discussions with the NRC Staff, and of the documents referenced.

We have seen preliminary drafts of the NRC Staff's report on lessons' learned from the Chernobyl Nuclear Station accident.

We agree that a thorough evaluation of the accident should be performed and that the lessons learned should be applied in the NRC regulatory process.

Although we have not reviewed the Staff's report in detail (nor have we seen a final draft), we consider the proposals made in the drafts we have seen to be sound and have no suggestion for any major changes in direction.

Sincerely,

.b b William Kerr Chairman References.

1.

Memorandum dated December 5,1986 from Themis P. Speis, NRC, to Raymond Fraley, ACRS,

Subject:

Chernobyl Information for ACRS Review, with y

enclosure:

revised draft of the Chernobyl Implications Assessment Report dated December 4, 1986 2.

Memorandum dated October 30, 1986 from Themis P. Speis, NRC, to Raymond Fraley, ACRS

Subject:

Chernobyl Information for.Subcomittee Review, with enclosur,c:

Draft Chernobyl Implications Assessment:

Assessments of Candidate Issues dated October 30, 1986 13

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UNITED STATES o,,

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NUCLEAR REGULATORY COMMISSION o

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WASHINGTON, D. C. 20555 January 15, 1987 The Honorable Lando W.. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555

SUBJECT:

ACRS COMMENTS ON THE IMPLICATIONS OF THE ACCIDENT AT THE CHERN0BYL NUCLEAR STATION UNIT 4 During the 321st meeting of the ACRS, January 8-10, 1987, we considered the implications of the accident at the Chernobyl nuclear station as it relates to nuclear power plants in the United States.

This subject was also considered during our 320th meeting, December 11-13, 1986 and our 319th meeting, November 6-8, 1986.

In our review, we also had the benefit of meetings of our Subcomittee on Safety Philosophy, Tech-nology, and Criteria held on November' 5 and December 10, 1986, and discussions with the NRC Staff.

The Chernobyl accident reminds us that, although a large nuclear power plant accident somewhere in the United States is unlikely, it is not impossible.

We believe it is essential that a thorough evaluation of the Chernobyl accident be perfomed and any important lessons from this evaluation are used in evaluating the risk posed by domestic nuclear power plants.

We recognize that the NRC Staff has such a program under.

way.

We believe that the most important lesson to be learned from the Cher-nobyl accident is that high priority must be given to ensuring that the management and the operating staff of each plant are. competent and are motivated to operate the plant safely and in strict compliance with plant administrative controls.

Strong emphasis should be given to the adequacy of the training and to the ability of the responsible personnel to prevent, to manage, and to mitigate severe accidents.

The operating staff should include on-site personnel with engineering capability who fully understand the design and operating characteristics of the plant l

and the implications for plant safety.

Such a staff should know the l

basis for the engineering and safety decisions made during plant design.

Although these recomendations are not new, the Chernobyl accident has reemphasized their importance.

Chernobyl also reinforces the known importance of determining the extent to which containments are capable of dealing with accidents more severe than the currently specified "design bas.is accidents."

We recomend that the NRC Staff give continued high priority to its current effort to 15

The Honorable Lando W. Zech, Jr.

-2 January 15, 1987 f

examine the containment performance expected for operating nuclear power i

plants-and to examine improvements needed to ensure that risk is limited to an appropriate level.

Reactivity transients severe enough to damage a light-water-reactor core can be hypothesized.

Risk estimates, operating experience, and infoned opinion all indicate that such transients are very unlikely.

However, such estimates and opinions depend in part upon assumptions that person-nel will comply with the administrative controls for operation, rather than depending entirely upon inherent characteristics of the hardware and processes. Present methods.of risk assessment do not satisfactorily account for personnel errors of the sort that could lead to noncompli-ance with such administrative controls.

Operating experience cannot be extensive enough to give high assurance that such errors are incredible.

For these reasons, there should be a systematic reexamination of the potential for severe reactivity transients, with emphasis on the impact of human error.

Multiple rod ejection, cold water insertions, void collapse, boron depletion, inappropriate bypassing of exposed safety circuits, and the importance of positive temperature coefficients during early core life are examples of the events and conditions that should be restudied.

The levels of defense against severe reactivity transients should be identified and, if possible, appropriately codified.

Emergency response following the Chernobyl accident confimed the need to ensure that the Protective Action Guides developed for application in the United States are comparable with those in neighboring countries and the need to reexamine the national policy on the storage and use of radioprophylactic agents.

Since potassium iodide was administered to thousands of people in the Soviet Union as a result of the Chernobyl accident, we hope that useful drta regarding its health effects will now become available.

I Other emergency response items highlighted by the accident include the importance of effective procedures for relocating large population I

groups, protecting ground and other drinking water upplies, dtcr>ntam-inating land and facilities, and protective measures for minimizing radionuclide intake through food and other pathways.

The accident at Chernobyl reinforces a previous ACRS concern that the effects of an accident involving a large release of radioactive mate-rials outside containment might negate safe habitation of the control room and other necessary facilities of the affected plant, or other units at a multiple-unit site.

Sincerely, William Kerr Chaiman 16 u

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION n

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 9

0, WASHINGTON, D. C. 20555

• 1. ,g May 13, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Commission Washington, D.C.

20555

SUBJECT:

ACRS REPORT ON PROPOSED RESEARCH TO REDUCE SOURCE TERM UNCERTAINTY During the 325th meeting of the Advisory Committee on Reactor Safe-guards, May 7-9, 1987, we discussed a proposed research program for resolution of source term uncertainty areas as described in SECY 86-369, "Plan To Address Source Tenn Technical Uncertainty Areas." We also con-sidered BNL report NUREG/CR-4883, an evaluation of this program by panels of experts sponsored by NRC.

The ACRS Subcomittee on Severe Accidents considered this matter during a meeting on April 22, 1987 In our review, we had the benefit of discussions with the NRC Staff and the documents referenced.

We commend the expert panels. for their expedited review and for their comments concerning some very complex phenomena.

We agree generally with their findings and recommend that ths Staff give careful consid-eration to their suggestions in planning the proposed research program.

We make the following additional observations:

(1)

In our report dated June 10, 1986 in which we commented on NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Terms," we recommended that the Staff attempt to quantify the uncertainties that were identified.

The expert panels also noted that there are no quantitative estimates of the magnitude of the identified uncertainties.

We agree with the. panels that those planning the research programs need guidance as to which contribu-tors to uncertainty are most important.

To provide this guidance, the Staff should attempt not only to specify uncertainties in the descriptions of particular phenomena, but should also estimate their contribution to risk.

There is also a need for an estimate of the level of uncertainty that is acceptable in making regulatory decisions.

Although SECY 86-369 identifies areas of uncertainty, it does not indicate what luvel of uncertainty would be acceptable, 17

The Honorable Lando W. Zech, Jr. May 13, 1987 1

nor does it indicate how likely it is that the proposed research will reduce the uncertainty to an acceptable. level.

(2)

In the areas of steam explosirms and hydrogen combustion, one of the panels recomended a reduction in research activities.

For steam explosions within the vessel that lead to early containment failure, the consensus is that the conditional prob' ability for such an event is very smell (0.01), and thus need not be considered fur-ther.

This panel s'urther concluded that hydrogen combustion is reasonably well unoerstood and that uncertainty in its understand-ing contributes relatively little uncertainty to estimates of j

source terms and risk.

However, significant uncertainties do remain in regard to the generation of hydrogen during an accident.

With the evidence now available to us, we agree with the panel's recomendation.

(3) A panel concluded that information needed to reduce the uncertainty in risk estimates for direct containment heating (DCH) will not be available within the next four or five years, even if a crash I

orogram is implemented.

In light of this estimate, the panel recommended the exploration of plant changes (hardware or proce-dures) which would eliminate the secuence.

The panel also recon-mended that the DCH experimental program be reorganized to show the effects of water and structural failure on DCH.

We concur in both recorrendations.

In general, we conclude'that the existing program is too narrowly focused.

The program should be redirected to encompass a broader range of possible scenarios, including esti-mates of realistic mass flows from the vessel and possible vessel failure modes.

The question of what is credible in the various situations must not be submerged in some large computer code, but should initially be sorted out by more straightforward and trans-parent physical arguments concerning the range of possibilities.

(4) There has been considerable discussion of the uncertainty associ-ated with the chemical form of iodine, either volatile (eleruntal) or non-volatile (chemically bound as in CsI).

After the TMI-2 accident, the absence of elemental iodine led some to conclude that the estimated risk should be reduced by a factor of as much as 100 from risks reported in WASH-1400, where it was assumed that all of the iodine was in elemental form.

It is now reported that in studies conducted in the preparation of NUREG-0956, the difference in risk for volatile vs. non-volatile iodine is only about a factor of 3.

A lesser priority should be assigned to research in this area.

(5) We observe that estimates of accident progression at key points in the core melt sequence depend on the prediction, using inadeouately based computer codes, of such parameters as melt temperature and 18

The Honorable Lando W. Zech, Jr. May 13, 1987 time required for vessel melt-through. There appear to be signifi-cant uncertainties in the predictions of a number of these key parameters that tend to be masked by the codes.

Since vessel penetration, core-concrete interactions, and the concurrent release of fission products, for example, are all very sensitive to melt temperature, we urge that efforts, including both experiments and independent calculations, be made to provide some independent and more transparent assessment of the behavior of key parameters.

Comparison with another code. embodying the same underlying as-sumptions is not sufficient.

(6)

In light of the importance of containment behavior in determining the magnitude of the source term, we recomend that more attention be given to the identification and evaluation of other scenarius having the potential for leading to a large release of radioactive material.

Additional coments by ACRS Member Glenn A. Reed are presented below.

Sincerely, William Kerr Chainnan Additional Comments by ACRS Member Glenn A. Reed While I agree with the ACRS letter to reduce the research in described areas, I wish to focus on the panels' observation made as a first suggestion" in the general conclusions that a prevention technique of "depressurization" (procedures and design) was important "to make the problem go away."

I recommend that research be increased and accelerated on the depres-surization idea and that the research include application of depres-surization as an alternative technique for core decay heat removal.

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References:

1.

U.S. Nuclear Regulatory Commission Staff Document, "Plan to Address 4

Source' Term Technical Uncertainty Areas "

SECY-86-369, dated December 11, 1986 1

2.

Brookhaven National Laboratory Report, "Review'of Research on Uncertainties in Estimates of Source Terms From Severe Accidents in Nuclear Power Plants," NUREG/CR-4883, dated April 1987 i

3.

U.S. Nuclear Regulatory Comission Report, "Reassessment of the Technical Bases for Estimating Source Terms," NUREG-0956, Draft i

Report for Coment, dated July 1985 4

U.S. Nuclear Regulatory Commission Report, "Reactor Safety Study --

An Assessment of Accident Risks in U.S. Conaercial Nuclear Power Plants,"WASH-1400(NUREG-75/014),datedOctober1975 5.

National Research Council Report, "Technical Aspects of-Hydrogen Control' and Combustion in Severe Light-Pater Reactor Accidents "

dated 1987 6.

U.S. Nuclear Regulatory Commission Report, "A Review of the' Current Understanding of the Potential for Containment failure from In-Vessel Steam Explosions," NUREG-1116, dated June 1985, i

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UNITED STATES

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June 9, 1987 The Honorable Lando W Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS REPORT ON PROPOSED GENERIC LETTER ON INDIVIDUAL PLANT EXAMINATIONS FOR SEVERE ACCIDENT VULNERABILITIES During the 326th meeting of the Advisory Comittee on Reactor Safe-guards, June 4-6, 1987, we discussed a draft Generic Letter prepared by the NRC Staff as guidance for individual plant examinations (IPEs) for snere accident vulnerabilities.

The IPEs are a p'rt of an implementation plan for the Severe Accident Policy Statement.

The ACRS Subcomittee on Severe Accidents considered this matter during meetings on December 19, 1986 and on May 28, 1987.

In our review, we had the benefit of discussions with the NRC Staff and with represen-tatives of the Industry Degraded Core Rulemaking (IDCOR) Program. We also had the benefit of the documents referenced.

The letter in its final fonn, accompanied by a panoply of supporting documents, is intended to provide guidance to nuclear power plant lica-sees in their performance of the individual piant examinations referr(d to in the Comission's Severe Accident Policy Statement (50 FR 3213.', August 8,

1985).

Specifically, the Policy Statement states:

Accordingly, when NRC and Industry interactions on severe accident issues have progressed sufficiently to define the methods of analysis, the Comission plans to formu-late an integrated systematic approach to an examination 1

of each nuclear power plant now operating or under j

construction for possible significant risk contributors (sometimes called "outliers") that might be plant speci-fic and might be missed absent a systematic search.

The NRC Staff finds that the following five options could satisfy the examination requirements, if appropriately supplemented:

(1) A PRA may be utilized, provided it is at least at Level II or Level III and it uses current methods and data.

21 i

L

The Honorable Lando W. Zech, Jr. June 9, 1987 (2) The IDCOR Individual Plant Evaluation Methodologies (IPEMs) may be used, provided the enhancements in the NRC Staff evaluation are applied.

(The NRC Staff evaluation of the applicable IDCOR IPEH is included as an attachment to the generic letter.)

(3) A Level I PRA supplemented by an appropriat'ely evaluated source tenn method may be applied.

(4) A simplified PRA which uses reduced systems models for the core damage analysis and sequence grouping for the containment performance analysia may be applied with an appropriate NRC approval.

(5) Another systematic examination method may be applied with prior 4

NRC approval.

The NRC Staff requests documentation of the examination results, as follows:

(1) Certification that an IPE has been corapleted and documented as requested by the provisions contained in the generic letter.

(2) A listir.g of the dominant sequences leading either to core damage or to significant releases from containment and their i

frequencies for the plant, together with the screening criteria used to identify the sequences.

(3)

Identification and listing of the main drivers, or leading contributors, to the predicted core damage frequency.

(4)

Identification and listing of the main contributors to any predicted containment failure.

(5) A discussion of the potential areas of improvement identified in the plant examination which could reduce either the probability of severe accidents or the probability of large releases from severe accidents.

(6) A list of the most cost-effective potential improvements, including hardware changes as well as changes in procedures and training programs.

(7) An evaluation of the most promising improvements, disposition of those improvements, and an implementation schedule.

(8) Consistent with the assumptions made in the IPE, a description of organizational responsibilities related to severe accidents together with the steps taken to assure that personnel are 22

The Honorable Lando W. Zech, Jr. June S',

1987 properly trained, appropriate procedures are in place, and diagnostic instruments and essential equipment will be available and will function where needed.

The PRA methods are relatively well specified from recent exper,encr.

at least for internal events up to the core damage stage.

C th regard to the IDC0'l IPEMs, the NRC Staff has provided evaluations which lead to a large number of reconnended modifications. and ad-ditions that are needed to make the IDCOR IPEM option acceptable.

The ACRS generally supports the NRC Staff's evaluations of the IDCOR IPEMs.

We recognize that formulating guidance for an individual plant examination is a fonnidable task.

We comend the NRC Staff for the progress that has been made, and for the cooperation with industry, through the IDCOR program, that has produced a significant contribu-tion to the effort.

However, we believe that the proposed guidan:e that has been prepared is deficient in a number cf areas, and that unless it is improved before licensees are required to design a program and perform an examination, a number of important objectives of the program are unlikely to be achieved.

The suggested aoproach to plant analysis is divided into two segments called "front end" (i.e., the descriptior, of an hypothesized sequence from initiation to the beginning of tavere core damage) and "back end" (i.e., from the onset of severe core damage to release of radio-active material from containment).

The guidance emphasizes that the two segments are not altogether independent.

However, because the onset of severe core damage or core melt has become something of a milestone in many PRAs, this is probably a reasonable division.

The guidance given for the front end analysis in the current draf t is much more detailed, and would be much easier for an inexperienced group to follow, than is the guidance for the back end which deals primarily with post-core-melt severe accident progression and con-tainment performance.

We believe that the guidance given, and the methods suggested, can provide a reasonable basis for a scarch for vulnerabilities ia the pre-core-melt or preventive part of postulated sequences.

However, the so-called guidance on certainment system perfomance analysis, especially that part that deals with PWRs, appears to be a rather hurriedly assembled discussion of some of the problems and uncertainties likely to be encountered in the analysis of containment performance, with very little guidance on how to perform a search for vulnerabilities.

We recognize and support the NPC Staff's effort not to be overly pre-scriptive.

Furthermore, the contrast between the guidance given for the front end and the back end analyses reflects, to some extent, the relative state of development of information needed to perform an analysis of reactor system performance, compared to that needed to i

23

The Honorable 1.anoo W. Zech, Jr. June 9, 1987 describe containment ystem performance.

Nevertheless, it is our judgment that if licensees, especially those with limited PRA expe-rience, are faced with guidance on containment performance analysis as ambiguous as that in the current draft, they will be so mystified that they will have no recourse but to retain an outside group to carry out the analysis.

They will thereby miss one of the more important benefits of the IPE, that of becoming familiar enough with system per#crmance to be able to recognize vulnerabilities in their plants, and of becoming aware of expected system performance in a severe accident.

The guidance on containment analysis should bo improved before the letter is released.

We also believe that not enough guidance is given as to goals and objectives of the examination.

The draf t letter, in aescribing the Connission's Policy Statement, identifies the "overall goals of the policy" as "(1) to reduce the probability of a severe accident, and (2) should a severe accident occur, to mitigate, to the extent possible, its consequences to the public."

It identifies the purpose of the examination as providing "the basis for a utility's appreci-ation of severe accident behavior, recognition of the role of preven-tion and mitigation systems and procedures, and the development of an accidert.nanagement scheme." A licensee must also, having discovered possible vulnerabilities. identify potential areas for improvement, suggest corrective actions to achieve improvement, decide which improvements he thinks should be implemented (if any), discuss th3 decision not to make those Judged inappropriate, and give a schedule for effecting those changes that are planned: all of this before th-examination has been reviewed by the NRC 5taff. The licensee is alsu asked to develop an organized approach, including training to deal with many severe core damage accidents.

Vulnerabilities are not defined, either qualitatively or quantita-tively (except perha enced in the letter)ps by inference from some of the material refer-

, nor is there guidance as to the amount and kir,d of improvement that the NRC Staff will find acc(ptable.

The reason given for not providing further guidance is that there are no objec-tive standards, that each licensee must make a decision for himself as to the changes that are appropriate.

However, the reviewing NRC Staff will need to have some 'riteria to provide a basis for review.

It would save everyone a co....derable amount of thrashing about if more guidance could be given as to criteria to be used in detemining the adequacy of the IPEs.

From our discussions with the NRC Staff, we have concluded that the projected scope of the review described by the draft letter may be too ambitious.

Based on our earlier discussions with the NRC Staff, we had concluded that the IPEs were to be perfortned to look for "out-lier" plants, i.e. plants with features, procedures or other operat-ing characteristics which produced risks unexpectedly high compared 24

The Honorable t.ando Wo Zech, Jrc June 9, 1987 with those of the general population.

It appears, however, that the program currently envisioned is one which attempts to establish a profile of core melt frequency and containment perfonnance (described at least semi-quantitatively, if not quantitatively) for each operat-ing plant, and then (possibly) attempts to reduce plant risk to some unspecified icvel, not necessarily the same for each plant, by requiring plant or other modifications which reduce the contribution from some selected population of risk contributors.

It would also lead to the beginning of a risk management program at ea'h plant.

Although there may be merit in this approach, we question whether many of these tasks are suitable for individual initiatives; rather they would need the efforts of appropriate new owners' groups, and NRC Staff guidance would have to be improved.

The guidance provided makes it clear tb t analyses of severe accident sequences initiated by external events and by sabotage are not requested at this tine.

Analyses for external initiators will be required later.

Nevertheless, it would be helpful to give at least some guidance at this time as to what is likely to be asked for later on, especially since one option given a licensee is to perfonn a PRA which considers external events.

In light of both the difficulty and the importance of the IPEs, we recommend that instead of the approach proposed in the draft letter, which has all operating plants begin the review innediately, the NRC Staff arrange trial reviews of several plants to be carried out cooperatively with licensees in somewhat the same way that the Systematic Evaluation Program (SEP) reviews were performed. Although part of the the review process developed by IDCOR has been exercised by them on several plants, the NRC Staff's view is that IDCOR's treatment of containment perfonnance does not consider several important safety-related questions.

Furthennore, for most of the plants reviewed by IDCOR, a more extensive PRA existed. Such reviews provide a useful reference. However, it would be valuable to perform reviews for a few plants that do not have PRAs.

If this were done cooperatively by the NRC Staff and the licensees, it could provide additional infonnation on the application of non-PRA approaches, and could also serve as a tool for development of more sharply focused guidance for later IPEs.

Sincerely, William Kerr 1

Chairinan l

I 25 i

l The Honorable Lando W. Zech, Jr. June 9, 1987

References:

1.

Letter from T.

P.

Speis. NRC, to W.

Kerr, ACRS.

Subject:

Documentation Necessary for the Initiation of the Severe Acci-dent Policy implementation, dated May 15, 1987.

2.

Letter from T.

P.

Speis. NRC, to W.

Kerr, ACRS,

Subject:

Documentation Necessary for the Initiation of the Severe Acci-dent Policy implementation, dated May 22, 1987.

3.

U.S.

Nuclear Regulatory Comission, "NRC Policy on Future Reactor Designs - Decisions on Severe Accident Issues in Nuclear Power Plant Regulation," USNRC Report, NUREG-1070, dated July 4

1985.

4.

IDCOR Program Reports (IDCOR-IPEMs):

(a) Technical Report 86.3A1, "Individual Plant Evaluation j

Methodology for Pressurized Water Reactors," April 1987, (b) Technical Report 86.3A2, "1PE Source Term Methodology for PWRs," March 1987, (c) Technical Report 86.381, "Individual Plant Evaluation Methodology for Boiling Water Reactors," Volun;es I and II, April 1987, (d) Technical Report 86.3B2, "lPE Source Term Methodology for BWRs," March 1987.

5.

Brookhaven National Laboratory Reports (Draf t), "Prevention and Mitigation of Severe Accidents," NUREG/CR-4920, dated March 1987 a

Volume 1. "BWR, Mark I Design,"

b Volume 2. "BWR, Mark 11 Design,"

c Volume 3. "BWR, Mark 111 Design,"

1 d

Volume 4, "PWR, Laroe Volume Containment Design," and e

Volume 5. "PWR, Ice Condenser Containment Design."

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76

63 Ct 0%

o UNITED STATES p",.g [ g NUCLEAR REGULATORY COMMISSION g

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September 16, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON CODE SCALING, APPLICABILITY AND UNCERTAINTY METHODOLOGY FOR DETERMINATION OF UNCERTAINTY ASSOCIATED WITH THE USC 0F REALISTIC ECCS EVALUATION MODELS Juring the 329th meeting of the Advisory Committee on Reactor Safe-guards, September 10-12, 1987, we reviewed the methodology developed by the NRC Office of Nuclear Regulatory Research for determination of the overall uncertainty associated with the use of realistic models, includ-irg related computer codes, for the calculation of thermal-hydraulic phenomena associated with loss of coolant accidents (LOCAs).

In our review, we had the benefit of discussions with representatives of the Office of Nuclear Regulatory Research (RES) and the Office of Nuclear Reactor Regulation (NRR).

Subcommittee meetings during which this topic was discussed were held on April 29-30, 1986, August 28, 1986, April 29-30, 1987, and August 4, 1987.

We also had the benefit of the docu-ments referenced.

A recently proposed revision to the ECCS Rule (10 CFR 50.46 and Appendix K) will permit use of realistic or "best estimate" methods in demon-strating that a peak cladding temperature (PCT) of 2200*F will not be exceeded during a LOCA.

This is in contrast to the original version of the rule which insisted on the use of a number of conservative assump-tions which were believed to provide an overestimate of PCT large enough t

to account for uncertainties. With the new rule change, a licensee may demenstrate that the calculated PCT, when adjusted with an appropriate allowance for overall uncertainty, has an estimated 95% probability of not exceeding 2200*F.

In our September 16, 1986 letter to you coment-ing on the proposed ECCS Rule, we noted the following:

l "The acceptability of realistic evaluation models rests on the development of satisfactory methodology for determination of the overall uncertainty.

Most of the development work needed here is either ongoing or planned by the Office of Nuclear Regulatory l

Research.

We recomend that the methodology used to evaluate uncertainty be subjected to peer review. We also wish to review this work."

27

The Honorable Lando W. Zech, Jr.

-2 September 16, 1987 RES has developed a method for quantifying uncertainty in PCT which it refers to as the Code Scaling, Applicability and Uncertainty (CSAU) evaluation methodology.

CSAU is designed to address uncertairties in the capability of a code to extrapolate small-scale test data to full scale, to correctly assess a particular sequence of events, and to account for variability in important parameters.

The focus of CSAU is on the important thermal-hydraulic processes with detailed attention given only to those processes which contribute importantly to overall uncertainty.

The end product of the CSAU method is _ an estimate of the total uncertainty associated with the calculation of a key parameter (e.g., PCT) by a given realistic ther 'l-hydraulic code for a particular plant and a particular accident transier?.

It must be recognized that absent an abundance of full-scale LWR plant transient data, it is necessary to rely substantially on engineering judgement in lieu of a rigor' ous statistical analysis.

The CSAU meth-odology systematizes the application of this judgment for the derivation of a quantitative allowance for uncertainty.

We believe that the CSAU method proposed by RES offers an acceptable means to estimate uncertainty associated with the use of realistic codes.

However, we wish to note the following:

• The CSAU methodology has not yet been tested over a wide range of applications.

Currently, RES is in the process of demonstrating the applicability of the method by using it to detemine the uncer-tainties resulting from a large break LOCA calculation using the TRAC PF1/ MOD-1 code.

While it appears that CSAU will be success-fully applied to TRAC, we recomend that RES complete an adequate eva!uation before the methodology is judged acceptable for use in regulatory actions.

• Before CSAU can be applied to a given code, complett o.umentation (e.g., code manual, model and correlation quality assurance docu-ment, and assessment reports) is necessary.

In the past, such

. thorough docurtntation has not always been available for licensing codes.

We recomend that steps be taken to ensure that future development of codes for licensing activities be performed in a nanner that ennres completion and availability of needed docu-mentation before the code is released.

l l

• The codes used to analyze thental-hydraulic behavior are very large and complex.

Validity of calculated results is dependent on the competence of the code user and the way in which the code is used.

For CSAU to be effective, the code developers, assessort and users must use the code consistently.

We recomend the NRC Staff take the necessary steps to ensure that proper controls are established.

l 28

The Honorable Lando W. Zech, Jr. September 16, 1987

• In order to ensure the ultimate success of the method, we believe it is necessary for RES to direct its experimental thermal-hydraulic programs appropriately to the needs of CSAU.

These experimental programs include the MIST, 2D/3D, and ROSA-IV coopera-tive efforts.

• We wish to caution that use of the CSAU method for regulatory applications will require the maintenance of an ongoing high level of competence and experience on the part of the hRC Staff members.

We suggest that the NRR call upon RES for such support as neces-sary.

We are encouraged by the move toward the use of realistic calculations for ECCS/LOCA phenomena.

We intend to follow the progress of this effort closely, and we wish to be kept informed.

Additional comments by ACRS Member Harold W. Lewis are presented below.

Sincerely, William Kerr Chairman Additional Comments by ACRS Member Harold W. Lewis I support the Comittee's letter, but do wish to add some cautionary notes about the misuse of some familiar words, which can lead to po-tential misuse of the CSAU (so-called) methodology.

To begin with, I support the move to "realistic" evaluations, since I believe that all evaluations should be made as honestly and realistical-ly as possible, after which regulatory conservatism can be applied cleanly and openly.

That is the thrust of this effort, and is fine.

Unfortunately, however, the words "best estimate" are often used inter-changeably with "realistic" to describe calculational techniques, and that is an error.

To a statistician, a best estimate is an estimate taken from the top of a probability distributinn, and that is simply a different idea.

This is not sophistry, since the misunderstanding of words that have established technical meanings can lead to incot rect I

calculaticns. To call an apple an orange does not make it one.

We were also briefed about a set of calculations in which parameters and assumptions were varied to provide a feel for the sensitivity of the 29

The Honorable Lando W. Zech, Jr. September 16, 1987 results to the specific assumptions made.

That is a reasonable way to learn about the sensitivity, but is not a way to learn about the "uncer-tainty" in the result, as any statistician would understand the word uncertainty.

Statistical uncertainty in its simplest fom is based on the concept of random samoling from a population of known characteris-tics but unknown parameters.

In that case, one can learn the uncer-tainty in an estimate of 'a parameter by studying the variance in a set of measurements, but that is not the situation here, where the variance in the results bears no relation whatever to any uncertainty, in any credible statistical sense.

The only reason for saying this is that in the familiar case of a normal distribution of sample measurements, one cen estimate the uncertainty from the variance, and thereby estinate the probability that the mean of the measurements differs from the true value by any ratio.

One can also estimate "confidence levels," but that is another saga.

None of that is true here, and this is again not sophistry.

In particu-lar, the draft Regulatory Guide supporting the proposed rule has state-ments about the "95% probability limit " "confidence level " and such things, and even states that the "use of two stcndard deviations for evaluating the 95% probability level is acceptable."

None of this is possible within the framework described, and simply reflects confusion on the part of the Staff about fundamental statistics' concepts.

I still support the letter and the program, since it is a major step forward, but repeat a recomendation I have made many times:

the NRC would benefit greatly by hiring a few good statisticians. One cannot do competent safety analysis in the presence of uncertainty (popular use of the word) witnout doing the statistics carefully.

References:

1.

U.S. Nuclear Regulatory Comission, Proposed Rule, "Emergency Core Cooling Systems, Revisions to Acceptance Criteria," February 26, 1987.

2.

U.S. Nuclear Regulatory Comission, "Request for Coments on Draft Regulatory Guide, ' Be s :. Estimate Calculations of Emergency Core Cooling System Performance,'" March 1987.

3.

U.S. Nuclear Regulatory Comission, NUREG-1230, "Compendium of ECCS Research for Realistic LOCA Analysis," April 1987.

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.E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

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e WASHINGTON, D. C. 20655 September 16, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON DEVELOPMENTS IN EMERGENCY PLANNING During the 329th meeting of the ACRS, September 10-12, 1987, we met with representatives of the New York Power Authority, the Oak Ridge Associ-ated Universities, and the Sandia National Laboratories to discuss preliminary analyses of the benefits of various measures taken to protect the population in case of a major accident at a nuclear power plant.

Representatives of the NRC Staff took part in the discussion.

This matter was also the subject of a meeting of our Subcommittee on Occupational and Environmental Radiation Protection Systems held on June 22-23, 1987 Studies reported by these groups indicate that sheltering, followed by monitoring of radiation exposure rates and relocation of populations from affected high radiation areas, within 4 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after an acci-i dent, yields predictions-for the number of prompt fatalities lower than those estimated to be provided'by the evacuation expected under current decision-making practices.

This was the statistical result of a wide range of accident scenarios.

Since the number of people, the distance they would need to be moved, and the disruptive impact of the sheltering-relocation approach would normally be less than those for the immediate evacuation approach, we believe that the NRC Staff should be asked to conduct an independent and prompt assessment of these findings.

Should this assessment confirm the reported observations, there appears to be reason for emphasizing sheltering, where appropriate, in nuclear emer-gency response.

Sincerely, William Kerr i

Chairman 31

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February 10, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U. S. Nuclear Regulatory Commission Washington, D. C.

20555

SUBJECT:

ACRS COMMENTS ON TESTING OF CHARCOAL ADSORPTION CAPACITY During the 322nd meeting of the Advisory Comittee on Reactor Safe-guards, Febrt:a, y 5-7, 1987, we discussed the capability for testing charcoal adsorption capacity in filters used at nuclear power plants.

Current Technical Specifications require periodic testing of the char-coal in adsorption units designed to control releases of airborne radioiodine from nuclear power plants.

However. "round robin" tests have shown that most commercial laboratories, both in the U.

S.

and abroad, lack the capability to determine the adsorption capacities of filter charcoals on an accurate and reliable basis.

Although the NRC has supported research on this problem, current NRC plans are to terminate this support, based on the expectation that industry will assume responsibility for continuing this research.

In connection with NRC termination of this work, we believe that the industry group that is to assume responsibility to continue this work should be identified and assurances made that the program will be pursued to a successful completion.

Sincerely, h1 William Kerr Chaiman l

l 33

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WASHINGTON, D. C. 20559 July 16, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Connission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON RESEARCH INTO CONTINUOUS CONTAIP'ENT LE',K/.6 MONITORING In some of our recent discussions, the concept of continuous containment leakage monitoring has resurfaced.

We believe there may be merit in this concept for reducing the risk of exposure to the public and plant operators in severe accident situations.

We recommend an investigation of continuous containment leakage monitor-ing to see if it can be helpful in risk reduction and if it is cost beneficial.

Sincerely, William Kerr l

Chairman 1

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5 NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O

WASHINGTON, D. C. 20655

... 4 August 10~ 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

PRELIMINARY ACRS VIEWS ON FIRE RISK RESEARCH SCOPING STUDY During the 328th meeting of the ACRS, August 6-8, 1987, we discussed the scope, direction, and current status of the Fire Risk Scoping Study being performed by Sandia Nation 61 Laboratories (SNL) for the NRC.

The ACRS Subcommittee on Auxiliary Systems also discussed this matter at a meeting on July 23, 1987.

In our review, we had the benefit of discus-sions with representatives of the Office of Nuclear Regulatory Research (RES) and SNL.

In the ACRS report of February 19, 1986 to the Congress on the FY 1987 NRC safety research program, and also in its June 11, 1986 report to the Commission on the FY 1989 research program and budget, the Comittee expressed concern about terminating the fire protection research program at the end of FY 1986 and recommended that funding for research in this area be restored.

The RES response to this concern was the initiation of a scoping study on the risk of fires to determine if further fire-related research is warranted.

This study is to utilize results of completed research and the fire risk analysis which is now nearing completion for the LaSalle County Station nuclear plant.

In the ACRS letter of July 16, 1986, the Committee expressed concern about the loss of program information and momentum that would refult from premature termination of ongoing fire-related research activities while awaiting the results of the scoping study.

The Comittee noted that termination of the needed research would be a serious loss, and would be costly to reconstitute later.

l Althcugh the Commission agreed with the ACRS on the importance of fire i

protection research, it did not restore the funding.

However, it did direct the Staff to work closely with the Committee to assess.further l

1 i

37 1

1

The Honorable Lando W. Zech., Jr. August 10, 1987 research needs and to consider the priority that should be assigned to fire protection research.

A good relationship was established and efforts are proceeding on schedule to assign a priority to possible research needs.

Various tasks are now progressing and the work is scheduled for com-pletion in December 1987.

The study includes identification of various potential fire-related issues, including t;.ose cited by the ACRS, and an assessment of the risk significance of such issues.

The risk consid-erations include an assessment of uncertainties in various previous probabilistic risk assessments (PRAs) and a requentification of PRA fire scenarios.

The final task will deal with the completeness of 10 CFR Part 50, Appendix R and other fire-associated regulatory requirements as they may relate to potential fire issues. Although the scoping study is still under way, we believe that the Comission may wish to be infonned of our preliminary views which follow.

j The main objective of the Fire Risk Scoping Study is to assess the risk significance and dominant sources of uncertainty associated with fire risk issues, with a final goal of assigning an appropriate priority for possible fire-related research.

We believe that the study is progress-ing satisfactorily toward this goal and is targeting the various con-cerns expressed by the. Committee.

The scope appears to be providing a needed and timely basis for determining priorities.

We plan to review and issue coments on the final results of the scoping study.

The redommended priority and the technical aspects of any proposed fire research program, including interim or long-range budgetary needs, will be discussed at that time.

Sincerely, William Kerr Chainnan 1

38

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UNITED STATES 8 ' '

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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g

WASHINGTON, D, C 20565 March 10, 1987 i

MEMORANDUM FOR:

Mr. Victor Stello, Jr.

E):ecutive Director for Operations FROM:

'Mr.

ey Executive Director, ACRS

SUBJECT:

ACRS COMMENT ON PRIORITIZATION OF GENERIC ISSUE 61:

"SRV DISCHARGE LINE BREAK INSIDE THE WETWELL AIR-SPACE OF BWP MARK I AND MARK II CONTAINMENTS" During the Comittees' review of the fourth group of generic issues at the 319th meeting of the ACRS (November 6-8,1986), the members deferred coment on the subject issue pending additional review. The Comittee has now completed review of this matter and its coments are attached.

As the Comittee has "agreed with coment" on Generic Issue 61, the members have requested a written response from the NRC Staff to their coments.

Please note that the attached coments recomend the evaluation of a potential new generic issue.

Attachment:

As Stated

(

39

Comment on Priority Ranking for Generic Issue 61 "Agrees With Comments" Generic Issue No:

61

Title:

SRV Discharge Line Break Inside the Wetwell Airspace of BWR Mark I and Mark II Containments Priority Ranking Proposed by the NRC Staff:

DROP ACRS Comments:

The ACRS agrees with the proposed priority ranking for this issue.

However, during our consideration of this issue, a related concern arose:

the issue of potential containment overpressuriza-tion given a steam and/or large coolant release in the drywell and bypass to the wetwell airspace through a stuck-o wetwell-to-drywell vacuum breaker (pen air-return valve). Our inquiries to NRP indicate that this particular accident scenario has never been directly addressed by the Staff. We recommend that NRR evaluate this item as a potential generic issue to ensure its appropriate resolu-tion..

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June 9, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

Dear Chairman Zech:

ACRS COMMENTS ON THE NRC STAFF PROPOSAL FOR THE R

SUBJECT:

USI A-44, "STATION BLACK 0UT" 4-6, 1987, and in our 32Sth During the 326th meeting of the ACRS, June 1987, we discussed the resolu meeting on May 7-9, We also "Station Blackout," that is being proposed by the NRC Staff.

discussed the Nuclear Utility Management and Resources Comittee initiatives directed at reducing the risk from "Station (NUMARC)

A Subcorraittee meeting was also held to discuss this issue Blackout."

During these meetings, we had the with the NRC Staff on May 6, 1987.

benefit of presentations by representatives of the NRC Staff and NUMARC.

We also had the benefit of the documents referenced.

members of the ACRS have considered and discussed 30, 1982, Since Merch nine meetings, and offered coments to the Executive this issue at 13, 1983 and March 12, Director for Operations in letters dated July The ACRS has been generally receptive to and supportive of the 1985.

Staff's efforts in seeking resolution of the issue.

We consider the proposed resolution of USI A-44, "Station Blackout," to However, we do be workable, and we corrvrend the f.taff for its efforts.

not recomend issuance of the final rule at this time.

J 1

We believe that the NUMARC initiatives may be a viable alternative for dealing with this issue on an expeditious schedule and may require the We believe least expenditure of resources on the part of the industry.

that the electric utility industry has a strong incentive to deal with t

"Station Blackout."

One shortcoming of the proposed NUMARC initiatives is the absence of a i

requiretrent for any assessment of a plant's ability to cope with station blackout for a specified length of time.

A letter from NUMARC has advised us that they are developing a methodology to do this, but that industry-wide agreement will have to be obtained.

They expect that the development of their initiatives will be substantially completed by l

September of this year.

41 l

l The Honorable Lando W. Zech, Jr. June 9, 1987 We recomend that the Staff continue to work with NUMARC on the techni-cal aspects of the NUMARC ef forts.

If by September of this year it is determined by the Staff that the NUMARC initiatives will not be effec-tive or timely in reducing the risk from "Station Blackout" to accept-able levels, or that the NUMARC initiatives will be unduly difficult to evaluate on a plant-to plant basis, we then recomend issuance of the final rule.

Additional remarks by ACRS Members Glenn A. Reed and Charles J. Wylie are presented below.

Sincerely, William Kerr Chairman Adoitional Remarks by ACRS Members Glenn A. Reed and Charles J. Wylie We believe the NRC Staff has done a comendable job in bringing A-44 to resolution.

However, we continue to support two previous ACRS letters (July 13, 1983 and March 12, 1985) recommending in part that A-44 implementation should be integrated with A-45, "Shutdown Decay Heat Removal Requirements."

Unfortunately A-45 has not arrived at the same status, and the NRC Staff wishes to proceed now with a rule and guide on station blackout which deal with A-44 only.

But, the root issue is not station blackout but rather decay heat removal to limit core melt risk to an appropriate level.

We do not consider it in the best interest of nuclear safety to proceed now with an NRC rule and guide on station blackout, which could compro-mise future desirable and more effective action for decay heat removal.

Sir,ce it appears that NUMARC-Nuclear Utilities Group on Station Blackou't (NUGSBO) has also been moving forward with an industry effort, and since the electric utilities should have premiere capabilities to upgrade vulnerabilities to station electrical blackout, we recomend NUMARC-NUGSB0 carry the ball, with NRC Staff interfacing and monitoring -- but without an NRC rule. This arrangement would leave the NPC uncompromised to act appropriately on A-45 when its resolution is completed.

In our opinion there may be sorte outlier units for which it is more preferable to focus and expend funds on the root issue of decay heat removal without diverting effort to station blackout; and such focusing may be more harronious with the backfit rule.

42

The Honorable Lando W. Zech, Jr. June 9, 1987

References:

1.

U.S. Nuclear Regulatory Commission, Federal Register Notice (51 FR 9829) for the proposed Station Blackout Rule (10 CFR 50.63),

published on March 21, 1986.

2..

U.S. Nuclear Regulatory Guide on "Station Blockout," dated March 30, 1987 3.

U.S. Nuclear Regulatory Commission, NUREG-1109, "Regulatory /Backfit Analysis for the Resolution of Unresolved Safety Issue A-44,"

submitted March 30, 1987.

4 U.S.

Nuclear Regulatory Comission, NUREG-1032, "Evaluation of Station Blackout Accidents at Nuclear Power Plants," draft, submit-ted April 16, 1987.

5.

U.S. Nuclear Regulatory Comission, NUREG/CR-3226 "Station Black-out Accident Analyses," dated May 1983.

6.

U.S. Nuclear Regulatory Comission, NUREG/CR-2989, "Reliability of Emergency AC Power Systems at Nuclear Power Plants, dated July 1983.

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o, WASHINGTON, D. C,20555

%4 October 15, 1987 1

The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555

SUBJECT:

ACkS COMMENTS ON THE PROPOSED RESOLUTION FOR GENERIC ISSUE 124, "AUXILIARY FEE 0 WATER SYSTEM RELIABILITY" During the 330th meeting of the Advisory Committee on Reactor Safeguards, October 8-10, 1987, we completed discussion of the status of a resolution for Generic Issue 124 (GI-124) concerning the reliability of auxiliary feedwater (AFW) systems in seven particular plants.

The Committee pre-viously(met with representatives of the NRC Office of Nuclear Reactor Regu-lation NRR) and the Office of Nuclear Regulatory Research during our 329th meeting, September 10-12, 1987.

This matter was also discussed during a meeting of the Decay Heat Removal Systems Subcomittee on August 5,1987.

We reviewed the beginning of this work about a year ago and comented in a letter dated September 17, 1986 to the Executive Director for Operations.

We also had the benefit of the documents referenced.

F GI-124 addresses concerns about the adequacy of AFW systems in a particular set of seven older PWR plants.

These plants had been singled out for generic attention in a screening study of AFV systein reliability several years ago.

It was believed that this group of plants deserved special attention in advance of the more general review of the reliability of decay i

heat removal (which includes the issue of AFW reliability) in all plants being evaluated in the Unresolved Safety Issue A-45 (USI A-45) program.

Each of the seven plants has a two-train AFW system estimated,4 at the time of the screening, to have an unreliability greater than 10' per demand l

Other "two-train" plants, which had estimated unreliabilities less than 10d per demand were not included in the group of seven plants.

Cur 1986 letter was critical of the proposed program plan because it failed to identify objective criteria by which reliability or effectiveness of AFW systems were to be judged.

The NRC Staff responded by asking that we wait l

until the initial plant reviews were available and then reconsider whether we agreed with their approach to resolution as put into practice.

We have now reviewed the initial plant evaluations and our objection to the l

process remains.

As we understand the resolution process, it is to consist of seven plant-specific evaluations and negotiated settlement packages, 45 Page Revised: 10/20/87

The Honorable Lando W. Zech, Jr. October 15, 1987 rather than a gereral solution.

Each evaluation starts with an inspection and review of the design and operation of a plant's AFV system by an NRR team.

The inspection and review identifies "negative features" in design, operation, or maintenance and calls these to the attention of the licensee.

It is then, apparently, the intent to correct or otherwise resolve these negative features to the mutual satisfaction of the licensee and NRR.

Our objection to this approach has twn main points:

(1) The quantitative criterion (unreliability creater than 10'4 per. demand) by which the seven plants were originally singled out as requiring special attention has been rejected by the NRC Staff as too crude" to be used in measurirg the adequacy of proposed AFW improvements.

This calls into question the original selection process.

It becomes unclear whether there really is a generic issue regerding AFW reliability in a certain subset of plants and, if there is., why these perticular seven plants are in the subset of concern.

(2) The NRC Staff has not specified an objective standard by which it intends to judge whether possible improvements to the AFW systems in these plants are adequate.

Instead, NRC Staff terms will review each AFW system in detail, react to what they find, and negotiate improve-ments with the licensees.

We believe this approach represents a serious misallocation of responsibility and resources between regula-tors and the regulated industry.

It is a mistake that should be corrected in this instance and in other regulatory activities as necessary.

We will expand on each of these two points below.

If the screening analysis used to identify this subset of seven plants as having a unique problem is now considered to be seriously flawed, then we believe the whole basis for GI-124 is invalid.

It may be most aparopriate to drop this issue and to concentrate Staff resources on the resolution of USI A-45.

If GI-124 is to be continued, the conditions important to AFW reliability should be considered more explicitly in the resolution.

From a risk per-spective, the minimum acceptable AFW reliability is related to the expected challenge or demand frequency on the system.

For example, if the main feedwater (MFW) system in a plant is capable of maintaining stable flow to the steam generators for an extended period following a reactor and turbine trip, then the reliability requirement on AFW might be lower than otherwise deemed acceptable for a plant without this capability.

Of course, if trips of the MFW system itself are a main cause of demand for Ani, this advantage might be unimportant. As another condition, if there is a strong capability for primary bleed and feed heat removal in a plant, again the reliability requirement on AFW might be lower than otherwise considered acceptable.

46

The Honorable Lando W. Zech, Jr. October 15, 1987 it appears to us that the plant reviews conducted so far have been done competently by experienced end capable review teams.

Negative features identified have been real and practical issues, but often of rather minor individual significance.

Some more significant design or operational problems have also been identified.

If all or most of the individual issues are corrected or improved, there is little doubt that AFW reliability will i

be somewhat improved at each of the plants.

This is a subjective judgment on our part because NRR has furnished no quantitative estimetes of the incrernental risk associated with each negative observation -- nor with their sum.

Our objection to this approach for resolution of GI-124 is not that the process itself entirely lacks merit, but that it is inappropriate for NRC to resolve a generic safety issue in this manner.

Inspection and review of the sort described to us should be carried out in-house by the utility-licensee or by an industry organization. The NRC should better use its own resources by providing the licensees with some objective def snition of the AFV relia-bility it believes is necessary.

For example, if an unreliebility for AFW greater than 10~4 per demand is judged by the NRC to be inconsistent with its overall intent in regulating nuclear power, then the resolution of GI-124 could require a good faith effort on the part of licensees to estimate the unreliability of the system in each plant.

This would be followed by licensee-initiated improvement of the AFW system sufficient to meet that requirement.

If the NPC believes analytical methods are not well enough developed to specify this sort of quantitative limit on unreliability, then it might in'trad want to specify a deteministic requirement, e.g.,

that two-train AFW systems are acceptable only if they incorporate certain favorable attributes or e diverse system for decay heat removal.

But, the NRC must then have the resolve to define these necessary ettributes in an understandable way and not resort to a reactive ("bring me a rock") style of regulation.

We recognize thet the dtvelopment of an appropriate objective criterion for i

AFW reliability is, or may be, a difficult task.

However, diversion of the i

engineering resources of NRP. to work that is more properly carried out by industry, such as the aforementioned inspection and review teams, only delays addressing the difficulty and may preclude development of a truly generic resolution that is both sound ard has long-tenn utility.

Sincerely, William Kerr Chairman l

47

L The Honorable Lando W. Zech, Jr. October 15, 1987

References:

1.

U.S. Nuclear Regulatory Coreission, "Safety Evaluation by the Office of Nuclear Reactor Re Auxiliary Feedvater System Reliability (Generic Issue 124)gulation, With Respect to Prairie Island Nuclear. Generating d

Plant Unit Nos. I and 2,"

transmitted by "letter from George lear, Division of PWR Licensing-A, Office of Nuclear Reactor Reoulation, to Dave M. Pusolf, Northern States Power, dated November 26, 1986, 2.

U.S. Nuclear Regulatory Connission, "Safety Evaluation of the Auxiliary s

Feedwater System (Generic Issue 1?A) With Respect to Arkansas Nuclear One Generating Plant Unit 2," transmitted by memorandum from Eric S.

Beckjord, Office of Nuclear Regulatory Research, to Thomas E. Murley, Office of fluclear Reactor Regulation, dated July 13, 1987.

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July 15, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COM1ENTS ON LICENSEE EVENT REPORTS PERTAINING TO CONTROL ROOM HABITABILITY For a number of years, members of the Advisory Committee on Reactor Safeguards, supported by ACRS Fellows, have examined Licensee Event Reports (LERs) pertaining to air cleaning, ventilatir,g, and monitoring systems at comercial nuclear power plants.

Several of the more recent of these studies have concentrated on LERs specifically pertaining to control room habitability.

1 The latest of these studies, which covered the three-year period frem 1984 through 1906, has revealed the following information:

1.

Or, an annual basis, from 3% to 8% of all LERs pertained to systems related to control room habitability.

For the three-year period, 7% of all LERs were in this category.

This represented a total of over 500 LERs.

2.

Of the LERs in this category, 61% were due to problems involving air monitors.

Of these, 55% were due to problems with radiation monitors and 29% were due to prnblems with chlorine runitors.

Most of these events were reported as LERs because malfunctions of the monitoring equipment led to actuations of the control room emergency ventilation system.

The large number of LERs in this category indicates a need ta address attention to their origin and the need for corrective action.

Such events almost er vtainly reflect a lack of reliability on the part of certain types of air monitoring equipment.

Several approaches may be useful in planning corrective action.

Al-though malfunctions of chlorine monitors account for a significant l

percentage of the cited LERs, data for the past several years indicate essentially no problems with these types of monitors at certain nuclear power plants.

It might be beneficial to detennine whether such monitors l

are in use in these plants and, if so, what type they are end how they are maintained and operated.

Such infonnation could be useful in resolving some of the problems observed at other plants.

i 49 i

--,y

-e-

Mr. Victor Stello, Jr. July 15, 1987 A second approach might be for the NRC Staff to consider encouraging all nuclear power plant licensees to ado?t the provisions of the current Standard Technical Specifications whLch specify a time limit within which a defective air monitor would have to be repaired and placed back into service.

Such a requirement would help make the management at all plants aware of the need to purchase and install reliable air monitoring i

equipment and to maintain it in proper working order.

1 Sincerely, William Kerr Chaiman i

1 1

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g WASHINGTON, D. C 20655 July 15, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON THE EMBRITTLEMENT OF STRUCTURAL STEEL Surveillance samples of steel used in the pressure vessel of the High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratcry re-cently have shown that the nil-dgetility transition (NDT) temperature of steel irradiated slowly at 120 F can rise much more rapidly with ex-posure to fast neutrons than would be expected from the available experimental work obtained in test reactors.

This appears to be due to two causes:

a flux rate effect (A lower fast neutron flux embrittles more than the same fluence accumulated at a much higher flux in test re-actors.)

0 the difference in tgmperature (550 F for comercial reactor pres-sure vessels vs. 120 F for the HFIR)

This has led to a significant shift in the NDT of the steel at a fast neutron fluence lower by roughly a factor of 20 than that predicted by the correlations used in the past for low temperature irradiations.

This acceleration is independent of the copper content of the material.

This suggests that steel structures outside the pressure vessel in comercial nuclear power plants may have embrittled where such behavior was not expected.

We believe it would be prudent for the NRC to do the following:

1.

Detemine if the brittle failure of any structural steel component near the outside of the primary pressure boundary would have safety significance.

2.

Detemine, using the low temperature irradiation data now available from test reactors, whether an increase in the fast neutron fluence by a factor of 10-100 would be predicted to give brittle behavior in these components.

3.

Implement a research program which would assemble bettr:r infoma-I tion on the rate of shift of the NDT of structural steels in 51

.2 Mr. Victor Stello, Jr. ~ July 15, 1987

[

1 i

comercial nuclear power plants at these lower rates and tempera-tures.

t a

i 4

Include consideration of the accelerated shift in NDT as part of 1

the evaluation of structures in the program on plant aging.

i.

Sincerely, William Kerr Chairman i

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'S, *... + #*g December 8, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON MEMORANDUM FROM VICTOR STELLO, JR., EDO, DATED OCTOBER 7, 1987 REGARDING THE EMBRITTLEFINT OF STRUC-TURAL STEEL We are concerned and perplexed by your memorandum of October 7 (refer-enced).

There you conclude that, "the neutron shield tanks and support structures do not appear to pose any safety problems. The embrittlement can be conservatively predicted as an increase in transition temperature of the steel of as much as 400*F."

You support your conclusion with the statement, "These structures are in compression, so even with a 0.2 g earthquake, the tensile stresses generated appear to be too low for fracture initiation."

Studies indicate that the highest risk of sudden pipe rupture in the primary system arises from the failure of supports of a major component.

We can see no reason to be sanguine about the safety of operating nuclear power plants with the largest, heaviest component in the primary system supported on a structure, parts of which are fully brittle. This is unsafe by any type of analysis.

The average stress may be compres-sive, but it isn't the average stress which would determine the failure l

of the structure.

These supports are welded structures so there are regions with tensile stresses as high as the yield stress. They operate in a temperature gradient so there will be thennal stresses which are tensile in the cold (less ductile) regions.

They are uninspected 50 we have no real idea of what kinds of flaws are present, and flaw size is critical in any meaningful failure analysis, i

It would be imprudent to operate nuclear power plants with brittle structures supporting the pressure vessels.

We recomend that an early effort be made to gain answers to the following questions:

1)

Is the terperature of the support structure of the reactor pressure vessel in any operating plant now below its nil ductility transi-tion temperature (NDTT)?

2)

Will the temperature of the support structure of the reactor i

pressure vessel in any operating plant drcp below its NDTT before the plant's license expires?

53

i I

Mr. Victor Stello, Jr. December 8, 1987 We hope and suspect that the answer to the first question is "no."

However, it is not clear that we know this with any certainty.

The research program mentioned in your meinorandum is necessary and desir-able, but it is not clear that it will answer the safety-related ques-tions noted above in a timely manner.

Sincerely, i

William Kerr Chairman

Reference:

Memorandum from Victor Stello, Jr., EDO, to William Kerr, ACRS, dated

October, 7,

1987

Subject:

ACRS Coments on the Embrittlement of l

Structural Steel f

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j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

msmorou, o. c. mis August 12, 1987 1

i The Honorable Lando W. Zech, Jr.

Chairman U. S. Nuclear Regulatory Comission i

Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON THE ADVANCE NOTICE OF PROPOSED RULEMAKING:

DEGREE REQUIREMENTS FOR SENIOR OPERATORS During the 328th meeting of the ACRS. August 6-8, 1987, and our 327th meeting, July 9-11, 1987, we discussed SECY-87-101, "Issues and Proposed Options Concerning Degree Requirements for Senior Operators," which was prepared in response to public coments t.1 the proposed rule. Meetings of our Subcomittee on Human Factors were also held on July 15,1986 and June 24, 1987 to discuss this issue with the NRC Staff.

During these meetings, we had the benefit of presentations by the NRC Staff as well as, representatives of the Westinghouse Electric, KMC, and Delian corpora-tions.

We also had the benefit of the documents referenced.

On May 31, 1986 the NRC published an Advance Notice of Proposed Rule-making (ANPRM) to require all applicants for a Senior Reactor Operator (SRO) license to possess a baccalaureate degree in engineerin0 or physical science after January 1,1991.

Two hundred letters of public coment were received in response to the ANPRM of which approximately 1

98% indicated opposition to the NRC's proposal.

i The nuclear utility industry and the NRC have endorsed a systems ap-proach to performance based training. At the heart of performance based training is a detailed Job and Task Analysis (JTA) which analyzes the many tasks that must be performed to carry out the various jobs of j

personnel filling positions in nuclear power plants, including the position of SRO.

The tasks are further anal to detennine the

)

various knowledges, skills, and abilities (KSAs)yzed that one must possess to perform the tasks.

The analysis continues further to determine whether the KSAs should be obtained through formal education or through specific training in the classroom, in the laboratory, at a simulator, or by self-study.

A number of JTAs have been perfonned by licensees as part of the conver-sion to perfonnance based training; analysis of these JTAs has not shown that a college degree is necessary for Senior Reactor Operators to 1

55 J

The Honorable Lando W. Zech, Jr.

2 August 12, 1987 perform the tasks of their jobs to ensure safety of plant operations. A Peer Advisory Panel appointed by the Comission came to the same conclu-sion in 1982 and recomended against a degree requirement for SR0s. We have not been informed of any technical rationale for requiring a degree for SR0s at nuclear power plants; we conclude, therefore, that a degree requiremant for all SR0s is primarily a policy issue.

We strongly support the concept of having engineering expertise on each shift.

The Comission's req'11.ement of a Shift Technical Advisor (STA) was a step in that direction.

Further, the Comission's provision of the option to combine the STA function with one of the SR0 positions was a step 'to encourage greater integration of the resulting engineering expertise into shift operations.

The Comittee endorsed both of these actions.

The NRC Staff ind4 cates that the percentage of SR0s with a baccalaureate degree in engineering or physical science has increased from 1, in 1980 to 28% in 1987.

We are informed that the primary reasons for considering requiring all SR0s in the future to have degrees is to enhance professionalism in reactor operations and to make it more likely that the higher management positions in nuclear utilities will be filled by individuals with plant operations experience.

We endorse these purported goals but question whether they will be realized through the proposed indirect approach of requiring degrees of all SR0s.

We believe there is a more direct approach to achieving these goals than through the proposed rulemaking.

We recomend that the Comission formulate more specifically its con-cerns and the goals it desires to achieve.

The Comission then should meet with appropriate licensee representatives (e.g., NUMARC) to convey the need for increased attention to the areas of concern. The NRC Staff and the licensees should then work to develop solutions, programs, and schedules for implementation of any changes from current practice deemed necessary. We realize that proposed rulemaking is one method to gener-ate sufficient attention to encourage licensee initiative; however, we believe a more direct and less adversarial approach is preferable when the proposed action is not driven by clearly identified public. safety Cohcerns.

In sumary, although the purported goals of the proposed rulemaking are laudable, we think that the depth of the concern about adverse effects of the proposed rule should be reconsidered; many of the coments were received from individuals who are knowledgeable about personnel consid-erations in the work place.

We recomend a more direct approach to identifying and addressing the Comission's concerns.

56

The Honorable Lando W. Zech, Jr.

3 August 12, 1987 Additional coments by ACRS member Glenn A. Reed are presented below.

Sincerely, William Kerr Chainnan Additional Comments by ACRS Member Glenn A. ( g I applaud the ACRS letter and wish to add further support to it.

As a person who earned a university engineering degree and one who held an NRC SR0 license, I am opposed to the degree requirement for SR0s, as in my opinion it 11 not needed from a job task analysis viewpoint, is not in the interest of licensed personnel morale, is not needed in the interest of best safety of operations, and would lessen the experience qualifications of SRO personnel.

I have found that a college degree in engineering or applicable science will probably ensure that an SRO candidate will have an acceptable enough intelligence quotient to be able to take on-site training.

However, there is no assurance from the college degree achievement that the SR0 candidate will have the even more important oualifications of mechanical comprehension, logical reasoning, and appropriate personality.

My thirty plus years of hiring and working with licensed operators has convinced me that acceptable perfonnance in a battery of aptitude tests (IQ, mechanical comprehension, logical reasoning, and personality traits), coupled with appropriate experience and training, will provide the best SR0 performers and peopic in overall shift charge.

My experi-i l

ence also has convinced me that the Shift Technical Advisor concept that was endorsed some years ago by the NRC can provide the best engineering support, and the best future promotional cross-fertilization into utility top management, and in+> the vendor design field.

l

References:

l

1. SECY-87-101, April 16, 1987, Issues and Proposed Options Concerning i

Degree Requirements for Senior Operators.

2. Federal Register, Vol. 51, No. 104, Page 19561 Friday, May 30, 1987, Advance Notice of Proposed Rulemaking, 10 CFR Parts 50 and 55 Degree Requirements for Senior Operators at Nuclear Power Plants.

3. Coments pertaining to the Advance Notice of Proposed Rulemaking -

Degree Requirements for Senior Operatnrs, KMC, Inc., September 29, 1986.

57

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5 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 o,

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June 10, 1987 Y

The Honcrable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

PROPOSED INTERNATIONAL WORXSHOP ON QUALITY IN DESIGN AND CONSTRUCTION OF NUCLEAR POWER PLANTS We believe that the present hiatus in licensing actions and the near completion of construction of the present generation of nuclear power plants provides an excellent opportunity for the NRC to reexamine the question of how best to achieve quality in design, manufacture, and construction.

Because of the great reliance placed by the NRC Staff on extensive and expensive formal quality assurance (QA) programs, it is important that we attempt to determine whether these QA programs deserve credit for actually achieving quality in the constructed plant.

It now is assumed that programatic or implementation deficiencies in the QA program are indicative of corresponding deficiencies in the quality of the plant.

Conversely, it is assumed that if there are no defi-ciencies found in the QA program, the plant quality will be very high.

Unfortunately, we have been unable to determine that either of these assumptions is valid.

Because other countries with substantial nuclear power programs seen to have achieved levels of quality even higher than we have in the United States, and because oost of those countries do not require or utilize the kind of quality assurance programs that we do, it would be informative to review and discuss with them their philosophies, procedures, and practices to achieve quality in their plants.

We recomend, therefore, that you consider the organization and sponsorship of an International Workshop on Quality in Design and Construction.

We would be happy to provide further thoughts on the agenda for such a workshop and would be willing to participate in the 1

planning to the extent that our resources permit.

Sincerely, j

L William Kerr Chairman j

i 59

[@ RECyI'o UNITE D STATES

- g NUCLEAR REGULATORY COMMISSION o

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v,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

g WASHINGTON, D. C. 20555 September 15, 1987 The Honorable Landn W. Zech, Jr.

Chairman U. S. Nuclear Regulatory Connission Washing'.or., D. C. 20555

SUBJECT:

ACRS COWENTS REGARDING PROPOSED INTERNATIONAL WORKSHOP ON QUALITY IN DESIGN AMO CONSTRUCTION OF NUCLEAR POWER PLAUTS In our June 10, 1987 letter to you we proposed an inte. national workshop cn quality in design and construction of nuclear power plants.

In a memorandum dated August 7,1987 (attached) Mr. Stello agreed with the importance of assessing the contribution of quality assurance (QA) programs in assuring quality in nuclear pnwer plants, but concluded that consideration of the proposed workshop should be deferred because the construction of new nuclear pcuer plants in the United States is nearing completion and because the NRC Staff is concentrating on improving nuclear power plant operations.

The fact that therc is a paese in the design and construction of new plants in the United States is one of the reasons we think that now is the tiiiie to evaluate the worth of existing regulations and associated i

regulatory guidance in attaining quality.

The time for evaluation and possible change of the regulations and associated guidance is not after i

new applications have been submitted in accordance with existing regula-tions.

We believe that the present hiatus in licensing actions and the near completion of construction of the present generation of nuclear power plants provides an excellent opportunity for the NRC to reexamine in an orderly fashion the question of how best to achieve quality in future plants.

Further, we believe that such reexamination should take place before the agency's and the industry's memory of past difficultie.i with current regulations and guidance has been blurred or lost.

Although the part played by quality assurance in operation of power plants and in licensing of waste repositories was not made explicit in our letter of June 10, 1987, it was not our intent to exclude these important questions.

In fact, we believe that the proposed interna-tional workshop on quality should specifically include discussion of these items.

We believe that there is much that could be gained from better understanding of how quality in plant operations is achieved in the United States and abroad.

61

The Honorable Lando W. Zech, Jr. September 15, 1987 We believe that our recommendation to the Comission to reexamine at thii time the question of how to achieve quality is consistent with the i

draft NRC Strategic Plan which specifically makes the assumption that quality problems can be expected in the future.

Further, we believe that the recommendation is consistent with the goal of the NRC which is stated in the draft strategic plan to "ensure that nuclear power plants under construction are designed and constructed properly and are ready for safe operation," as well as being consistent with the stated goal to "prepare for future reactor licensing activities."

We would be pleased to meet and discuss this matter in greater detail with you in the near future.

Sincerely, William Kerr Chairman

Attachment:

Memurandun from Victor Stello, Jr., EDO, to llill-f am Kerr, ACRS, dated August 7, 1987, "Proposed International Workshop on Quality in Design and Construction 7of Nuclear Power Plants" 9

62

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON, D. C. 205s5 k...*,/

August 7,1987 MEMORANDUM FOR:

William Kerr, Chairman Advisory Committee on Reactor Safeguards FROM:

Victor Stello, Jr.

Executive Director for Operations

SUBJECT:

PROPOSED INTERNATIONAL WORKSHOP ON QUALITY IN AND CONSTRUCTION OF NUCLEAR POWER PLANTS The purpose of this memorandum is to respond to your June 10, 1987, letter to Chairman Zech concerning a proposed international workshop on quality in design and construction.

I agree with the importance of assessing the contribution of the licensees' quality assurance (QA) programs in assuring quality in nuclear power plant design and construction.

In this regard, the staff has implemented several key initiatives:

readiness reviews, integrated design inspections (IDIs),

and construction appraisal team inspections (CATS).

As current construction of nuclear power plants in the United States nears completion, we are focusing primary staff attention on improving nuclear power plant operations by applying available staf f resources to initiatives that include performance-oriented QA inspections in lieu of the programmatic-type QA inspections, safety system functional inspections (SSFIs), and safety systems outage modification inspections (SSOMIs).

We are also continuing to ensure high quality design, engineering and construction in support of modifications to existing operating plants.

These types of efforts should improve our efforts to foster programs that provide a better assurance of quality for standardized plants.

Because of the dedication of the NRC staff to these initiatives, I feel that we should defer consideration of the propos?d workshop at this time.

I believe it would be more appropriate for the, Commission to reconsider such a proposal when a clearer picture emerges of the next generation of nuclear power plants in the United States.

V or Ste Jr Executive Director for Operations CONTACT:

Jack W. Roe, NRR 49-24803 63

a stouq'o, UNITED STATES NUCLEAR REGULATORY COMMISSION o

3

Y ADVISORY COMMITTEE ON REACTOR SAFEGUARDS I

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WASmNGTJN, D. C. 20555 July 15, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON INTERNATIONAL COOPERATION ON RESEARCH RELATED TO RADIATION PROTECTION During the 327th meeting of the ACRS, July 9-11, 1987, it was brought to our attention that members of the NRC Office of Nuclear Regulatory Research are attem,,tirg to develop a system for the coordination of research being conducted ;n various countries on the biological effects and control of ionizing radiation.

These efforts hold promise for assuring that key problems are effectively addressed and for reducing unnecessary duplication and the wasting of re-sources. We endorse these efforts and encourage their support.

Sincerely, William Kerr Chairman j

65 i

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November 10, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMEliTS ON THE DEVELOPMENT OF RAPIATION PROTECTION STANDARCS During the 331st meeting of the Advisory Comittee on Reactor Safe-guards, November 5-7, 1987, we met with Floyd L.

Galpin, Office of Radiation Programs, U.S.

Environmental Protection Agency (EPA), and Pebert E. Alexander, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission (NRC), to discuss current developments related to

)

radiation protection standards.

These discussions included reports on the efforts of (1) EPA to establish individual dose rates for members of the public that would be considered to be ?below regulatory concern" (BRC), and (2) an interagency committee, coordinated by EPA with NPC support, that is engaged in developing guidance for federal agencies on radiation protection of the public.

These topics were also subjects of discussion by our Waste Management Subcomittee during its meeting on October 15-16, 1987.

Current EPA efforts are being directed prinarily to developing limits on dose rates from low-level radioactive wastes, including the development of dose rates that are BRC, for members of the public.

Several proposals on this topic from outside organi7ations have been reviewed and endorsed by the EPA's Science Advisory Board.

As such, this work holds promise for alleviating some of the problems being encountered in the management and disposal of such wastes.

Although these efforts have revealed inconsistencies in existing radi-atior, protection standards (which will require considerable efforts to resolve), and although problems remain (such as clarifying distinctions in dose rates considered to be BRC and those considered to be de mini-mis), we are very encouraged by these activities.

They hold liiFomise, not only of providing a coherent system of radiation protection stand-ards, but also of placing the risks fron low radiation dose rates in better perspective.

67 l

l' l

2-November 10,' 1987 The Honorable Lando.W Zech, Jr.

For these reasons, we recomend that the NRC continue its support'of and ~

lend. encouragement to the work -of.-the. interagency comittee and the related efforts of~the EPA, Sincerely.

William Kerr Chairman j

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s 68

o UNITED STATES

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NUCLEAR REGULATORY COMMISSION o

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

g WASHINGTON, D. C. 20555 January 15, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, DC 20555

SUBJECT:

ACRS RECOMMENDATIONS ON IMPROVED SAFETY FOR FUTURE LIGHT WATER REACTOR PLANT DESIGN During the 321st meeting ~of the ACRS, January 8-10, 1987, we completed our discussion of improved safety requirements and objectives for future light water reactor power plants (LWRs). This discussion began during the 316th ACRS meeting, August 7-9, 1986.

The scope of our present comments is limited to nuclear power plant design.

Other factors, such as plant operation and management, are necessarily involved, but are beyond the scope of our present remarks.

The ACRS has on several previous occasions recommended that future LWRs should be designed to be safer than current LWRs. This is not to ignore the excellent safety record thus far of LWRs in the United States. We believe this increased safety can be achieved with reason-able economy because better technology is available today.

Improved concepts for plants and improved underst:nding of risks have been developed over a generation of experience in design, operation, and analysis.

But, not all of these concepts have been incorporated into the newest reported LWR designs.

We believe many of these concepts can be incorporated with acceptable effect on plant cost or operating efficiency.

With the expectation that future plants will be stan-I dardized, the next group of plants to be licensed will probably set the safety design philosophy, and even details of implementation, to be used in nuclear power plants for several decades.

The mean estimates of risk from generation of electricity by the use of nuclear energy are at least as low as those for generation by other methods.

However, the acceptability of these estimates is much affected by the large uncertainty associated with them.

A compelling reason for implementing improvements -- apart from the fact that improvements are possible -- is to reduce the uncertainty in the risk estimates.

Future plants should be able to survive a wider spectrum of off-normal challenges and mistreatments.

For example, normal operating systems should be forgiving of most operational errors and imperfections in maintenance.

Accident management and mitigation systems should be designed, not for a narrow set of design-basis accidents, but to 69

The Honorable Lando W. Zech, Jr. January 15, 1987 reasonably accomodate a broad range, variety, and time sequence of threats.

i Our recomendations are based on insights provided from quantitative risk analyses, lessons learned from operating experience, and continu-ing concerns.

In the sections that follow, we list and discuss a number of possible safety improvements.

Several of these overlap, and we do not expect that all of them should be implemented.

Rather, we offer them with the belief that each is worthy of serious consid-eration in connection with future designs.

1.

Dedicated and Protected Decay Heat Removal System (DHRS)

We recomend for consideration that future LWRs include a ded-icated, protected, redundant, 9 cay heat removal system having its own power, fuel, and water supply, with a capability for makeup, including coolant lost from very small LOCAs, and for recirculation from the containment sump. This system should have a large seismic capability such that its function is not threat-enej by earghquakes having an occurrence likelihood in the range 10 ' to 10-per year.

There should be similar protection and seismic capability for the primary system and all components whose specific function is required for proper operation of the dedicated decay heat removal system, as well as protection against fires, flooding, and adverse environmental effects. This system should be capable of actuation but not termination from the control room.

We list this item first because the provision of such a system would alleviate our concerns in several areas, including the following:

If the DHRS is protected against fire, internal or external quakes at the 10,hy an ingider or by terrorists, and earth-floods, sabotage to 10-probability level, the degree of protection required of other portions of the plant against such events could be relaxed in many instances.

In addition to the economies these reductions might lead to, we believe that they might lead to relaxation or removal of many of the impediments to access and flexibility of operation that are no's imposed by tecurity and fire control.

The loss of all sources of AC power, both off-site and on-site (station blackout), would be of less concern if a DHRS is provided.

However, vital DC power and certain vital cooling functions (such as cooling of primary pump seals in a PWR) now performed by using AC power would have tn be dealt with appropriately.

in some of the further recomendations that follow, we indicate that the identified needs would be

reduced, or perhaps 70

The Honorable Lando W. Zech, Jr. January 15, 1987 eliminated, if a dedicated, protected, decay heat removal system were provided.

l 2.

Safety Train Redundancy The general principle of "N+2" trains should be adopted for actiste, safety-related functions.

N is defined as the number of trains required to perform a necessary safety function.

N is equal to one if the train has 100% capacity to perform the,

function.

N is equal to two if each train has 50% capacity.

Thus, an "N+2 rule" would require three 100% trains, or four 50%

trains.

Each cf the N trains would have its own independent support systems.

Each train would be physically separate from the others, and diverse designs or equipment should be censidered if this can be shown to provide a significant safety advantage.

Exceptions to this general principle should be permitted for systems providing functions with low risk potential and for systems which can be demonstrated to be exceptionally robust and reliable.

The proposed high level of functional capacity could be used to impreve plant availability by use of Technical Specifications which permit one of the extra trains to be out of service for maintenance and testing for somewhat longer periods than is now the practice for the first train of redundancy.

3.

Design of Containment Systems The need to mitigate the consequences of certain severe accidents should be considered explicitly in the design of containment systems (structures, penetrations, sprays, vents, etc.).

The severe accident sequences to be considered should be those for which the mitigation provided by the containment systems is required to meet the Comission's proposed general perfonnance guideline that the overall mean frequency of a large release of radioactive materials to the environment from a nuclear power plart accident should normally be less than 1 in 1,000,000 per reactor year of operation.

The severe accident sequences that neect not be considered are those of sufficiently low probability that. the releases, unmitigated by specially designed containment sys". ems, will in the aggregate not exceed this objective.

4.

Protection Against Sabotage

)

We are not of one mind on the issue of the extent to which LWRs should be protected against the threat of damacing sabotage by terrorists and insiders.

On the one hand, there is reason to believe that certain design choices can lead to inherently better resistance against such a threat, even if these choices are not specifically directed 71

The Honorable Lando W. Zech, Jr. January 15, 1987 against sabotage.

For example, control rooms can be -positioned so they are away from the exterior ground level and protected from truck bombs by existing massive concrete structures.

Good physical separation of redundant sa fety trains may provide significant inherent protection.

Sore of us favor hardening, or separation, or other protection of mast vital functions such that they are relatively well protected against transportable explo-sives.

If includea in the original design, part of these changes should result in modest added cost or modest loss of other beneficial plant characteristics.

On the other hand, some of the members are not convinced there is reason to believe nuclear power plants are particularly attrac-tive targets for saboteurs.

If a terrorist aims to actually cause injury to large numbers of the public, there are far easier and more effective targets throughout the country.

Also, with 120 oportting plants [today's population) built to a lower level of sabotage protection and a new set of plants built to a higher level of sabotage protection, this discrepancy will surely be noted and taken into account by a terrorist in the selection of a specific target, if the aim is to cause physical harm to the public.

It appears to these members that the resources society allocates for defense against terrorism would be more effectively used in areas other than nuclear power.

In the case of the insider, the ACRS believes the threat is of low probability.

This should not, however, discourage prudent design features which could impede insider actions or reduce the likelihood of success.

5.

Fire Protection, Those responsible for cotJucting probabilistic risk assessments 1

(PRAs) have not been very successful in quantifying the risk from large fires involving essential reactor systems.

As a result, the real benefit of existing fire protection provisions and backfits remains uncertain.

We believe future LWRs should be designed so.that cold shutdown of the plant using safety-grade equipment can be accomplished quickly (within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) in the event of any single fire which may burn up to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Physical separation and protective barrier or compartment arrangements should include a reasonable accounting for the adverse effects of the spread of heat and the products of combustion beyond the fire zone, including "nsequential spurious actuation of fire mitiga-tion features and the resulting damage to safe shutdown equip-ment.

Ef re mitigation features should be designed to function properly, and not to spuriously actuate, during or after a seismic event.

If the plant has a DHRS es discusseo above, only those other portions of the plant vital to accomplishing safe shutdown would 72

The Honorable Lando W. Zech, Jr. January 15, 1987 need to be protected against fire consistent with the more stringent requirements listed above.

6.

AnticipatedTransientsWithoutScram(ATWS_},

We suggest that design features be introduced that would make an ATWS event a much less serious, if not a negligibly small con-tributor, to risk.

For PWRs this might involve some combination of increased negative moderator temperature coefficient of reactivity and increased pressure-relieving capability for the primary system.

For BWRs a partial contribution would be made by something approaching 100% relief capability in the event of turbine trip or main steam isolation valve closure.

We also suggest that the combination of control and safety systems be cxamined for reliability, as well as for testing and maintenance of the systems, to reduce the need for some of what are now considered to be safety-related scrams, as well as to reduce the number of spurious scrams.

7.

Systems Interactions Operating experience and reviews of existing nuclear power plants have provided evidence of unanticipated adverse interactions from supposedly separate systems.

These supposedly separate systems sometimes interact in unanticipated ways because they are depen-dent on comon support systems (such as power supplies, comon piping systems, etc.) or because they share the same or adjoining physical space.

Those people responsible for performing PRAs can successfully incorporate the effect of these interactions only if they are known and understood and if probabilities of occurrence can be established.

We believe that further effort is warranted to develop techniques and processes which can seek out and eliminate such interactions.

8.

Electric Power Systems We believe that the frequency of transients and spurious reactor scrams should be reduced by providing electric pcwer supplies that are less vulnerable to transmission network disturbances.

We recommend that General Design Criterion 17 be revised to require that the circuit which is provided to be immediately available to cope with a LOCA be the normal power supply to the plant auxiliarias and safety systems and be supplied continuously and unswitched from the low side of the main stepup transfomer during and throughout startup, operation, and shutdown of the nuclear generating unit.

We believe that the capability of a plant to cope with the loss of all off-site power can be improved.

For one thing, the proposed resolution of Unresolved Safety Issue A-44, Station Blackout, should be implemented in the design of future plants.

73 l

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The Honorable Lando W. Zech, Jr. January 15, 1987 For another, the reliability of en-site AC power sources can be enhanced by designing the nuclear system with sufficient steam bypass, feedwater inventory and make-up, and run back capability to sustain unit load rejection from 100% power and to run back to "house" electrical load, or by providing an additional, pref-erably diverse, standby electrical generating unit. The need for these features would be reduced if a DHRS is provided, as dis-cussed in Item 1 above.

9.

Probabilistic Seismic Design Important safety systems should be explicitly designed using probabilistic seismic design methodology to survive and function during and after severe seismic events.

Only survivability and i

those functions needed to bring to and hold the reactor at cold shutdown need be considered.

A DHRS such as discussed above would reduce the number of structures and systems requiring very stringent seismic design.

i 10.

Primary Pressure Boundary We recomend that the primary system pressure boundary be de-signed and fabricated to minimize the number of welds and opti-mize the ease of inspecting them.

i

11. Dedicated Systems and Sharing There should be minimum sharing of equipment, flow paths, and support facilities among nominally separate systems.

12. Control Room Protection for Severe Accidents Safe habitation of the control room and other necessary facil-ities should be ensured in the event of an accident that results in a large release of radioactive materials outside containment.

For multi-unit sites, this requirement applies to both the damaged unit and other units on the site.

Additional coments by ACRS Members H. Lewis, F. Remick, P. Shennon, and D. Ward are presented below.

Sincerely, William lierr Chaiman 74

The Honorable Lando W. Zech, Jr. January 15, 1987 Additional coments by ACRS Members H. Lewis, F. Remick, P. Shewmon, and D. Ward.

This is a camel of a letter, describing a camel of a reactor. We have no reason to doubt that each of the features recomended in the letter may improve safety, nor do we have any reason to believe that there are not better and more cost-effective alternati-ves.

This problem is compounded to the extreme by putting them all together.

The purpose of this letter is presumably to distill the Committee's observations and experience with the current generation of reactors, designed over the last few decades, and to put that experience to work in expressing a design philosophy for the next generation of reactors.

There is no hint of a philosophy, but _ instead a laundry list of improvements, all unanalyzed.

Though the Comittee has of ten recom-mended that the next generation be safer than the past, that recomen-dation has never been justified.

It may be right, but seems to be inconsistent with the Comission's Safety Goal Policy.

There is no doubt in our minds that, with new technology and years of experience, a new generation can be either safer at comparable cost and level of complication, or equally safe at lower cost and greater simplicity, and that choice is so fundamental that it is, in our view, not respon-sible for the Committee to opt for greater safety and greater com-plication without analysis or justification.

We believe one can learn from experience and that the next generation must inevitably be better than the past (and thereby safer), but we are uncomfortable about designing those reactors in comittee.

Additional coments by ACRS Member, David A. Ward.

I disagree with the Comittee's recommendation that future LWRs should include a dedicated, bunkered decay heat removal system.

In my opinion, the safety advantage from such a system is higbly uncertain and likely to be very slight or even negative.

The cost would be great and there would be added complexity in operations.

I believe added reliability offered by adoption of the N+2 principle with some diversity and separation of trains is adequate and preferable.

The promises of trade offs, e.g., relaxation of requirements on main-line systems, are phantoms.

A systematic study to determine what should be included in a bunkered system and whether there would indeed I

be important trade offs might be warranted at this time, but the Committee has not made such a study.

The recomendation is a hip shot.

The Comittee has elected not to makt recomendations relative to either of a pair of weaknesses in LWRs which I believe make them the object cf criticism from the proponents of new reactor concepts.

These are:

1) absence of a backup scram system and 2) the fact that every scram, real or spurious, becomes a challenge to the plant 75

The Honorable Lando W. Zech, Jr. January 15, 1987 l

because of the necessity for emergency feedwater.

I believe consid-eration should be given to ' development of an independent backup scram system for LWRs.

This would include not only independent sensor and I

control logic, but also an additional system of absorber rods or other material (possibly a liquid) to rapidly and reliably enter the core.

Further, I believe there should be consideration of a passive or continuously operating decay heat removal system so that a reactor scram will not be a challenge, but instead always be an unambiguous shift to a safer operating mcde.

Beyond these two specific points, I believe the best approach for the NRC to take in implementing safety improvements, such as those sug-gested in this letter, in LWRs of the future is to incorporate them into a revised set of General Design Criteria.

Although iteration with designers and licensees will be necessary, the improvement process will bast be served by establishing a clear new basis at the beginning.

In addition, I am concerned that the concept of quality assurance, as applied in the nuclear power industry, has not been successful.

I do not, of course, question the need for quality nor do I have major concerns about the quality of existing plants. However, I do question whether QA has had much to do with either. This might not be so trou-blesome except that QA as practiced is very expensive and uses re-sources that might better be spent in other activities, including more effective reactor safety programs.

I suggest that the present hir.tus in plant design and construction provides an opportunity for -the Comission to rethink i tt. commitment to the present concepts and practices of QA.

76

/

o UNITED STATES

[n NUCLEAR REGULATORY COMMISSION

,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS O

g W ASHINGT ON, D, C. 20$$5 July 15, 1987 4

The Honorable Lando W. Zech, Jr.

Chairfnan U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON IMPROVED SAFETY FOR FUTURE LIGHT WATER REACTORS During the 327th meeting of the ACRS, July 9-11, 1987, we discussed your request (see Reference) for the name of an existing plant that incorporates the desirable' features recomended for consideration in our letter to you dated January 15, 1987 on Improved Safety for Future Light Water Reactor Plant Design, To the best of our knowledge, we believe that most of those features are incorporated in the Federal Republic of Gemany KWU-Standard pVR plant designed,' licensed and constructed under the KONVOI process.

Examples are Isar 2. Emsland, and Neckarvestheim 2, which are near completion.

Additional comments by ACRS Member Harold W. Lewis are presented below.

Sincerely, William Kerr.

Chairman Additional Coments by ACRS Member Harold W. Lewis I am greatly concerned lest the Comission may have misunderstood the list in r

our earlier letter as a list of recomendations of features the Committee l

views as desirable for incorporation into future reactors It sheuld be emphasized that those were reco.nrrended for consideration only.

None have l

been sufficiently analyzed by the Comittee to justify a stronger interpre-l tation.

l J

77 1

6 2-July 15, 1987 The Honorable 1.ando W. Zech, Jr.

Reference:

Memoran'dum dated Ap'ril 22, 1987 to Raymond F. Fraley, ACRS, from John C.

Hoyle, Assistant Secretary,

Subject:

Staff Requirements - Periodic Meeting.

with Advisory Comittee on Reactor Safeguards 7

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g WAEMNGTON, D. C. 20555 September 15, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON IMPROVED SAFETY FOR FUTURE LIGHT WATER REACTORS During the 329th meeting of the ACRS, September 10-12, 1987, we discussed two recuests transmitted in the memorandum from John C. Hoyle, Assistant Secre-tary, NRC, to Raymond F. Fraley, Executive Director, ACRS, dated April 22, 1987 (see Reference).

The ACRS Subcommittee on Future Light Water Reactor Designs had previously discussed these requests during a meeting on Septem-ber 9, 1987.

The first request was that "the ACRS pursue its review of the experience and design features of some of the European plants." We intend to continue such a review and will keep the Commission informed of our findings as appropri-ate.

The second request was that the ACRS "address the feasibility, benefit, and cost effectiveness of selected and combined systems recommended in the Kerr to Chairman Zech letter dated January 15, 1987.

The review should include plant reliability, challenges, complexity, and burden on plant and mainte-nance personnel."

We believe that such a study clearly is desirable.

However, it would require consideration of many aspects of design other than safety and is beyond our capabilities and resources.

For these reasons, it is more appropriate as a task for the NRC Staff or a centractor.

We would be pleased to discuss this with you further.

Additional comments by ACRS Member Glenn A. Reed are presented on the follow-ing page, i

Sincerely, William Xerr Chainnan 79

The Honorable Lando W. Zech, Jr. September 15, 1987 Addition'l Comments by ACRS Member Glenn A. Reed a

their advanced LWR designs (ghouse Electric Electric Company and Westin As you know, both the General on the drawing Corporation have stated that boards) do indeed incorporate most or all of.the features mentioned in the ACRS letter of January 15, 1987. As should be realized there's a long path between the drawing board and a built operating reactor, and therefore I recommend that the NRC sponsor an in-depth study as a follow-on to USI TAP A-45 that addresses the most important recommenda tion of the ACRS January 15, 1987 letter, the recommendation on a dedicated decay heat removal system.

The follow-on study should address decay heat removal for future LWRs and the systems, diversity of systems, redundancy of components, and the other complex safety influencing aspects such as security and fire.

The operating reactor KONVOI should not be excluded from the stJdy.

It is my opinion that an in-depth study may reveal cost savings and inproved operating and emer-gency potential for future LWRs.

In particular, I feel that the use of a backup primary blowdown (dedicated) depressurization and decay heat removal systen for PWRs will provide improved operations, less operating burden, fewer security demands, and reduced core melt probabili,ty.

Reference:

Memorandum dated April 22, 1987 to Raymond F.

fraley, ACRS, from John C.

Hoyle, Assistant Secretary,

Subject:

Staff Requirements - Periodic Meeting with Advisory Committee on Reactor Safeguards 80

9 UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Q

WASHINGTON, D. C. 20555

'o.....*

April 13,1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS ACTION ON THE PROPOSED REGULATORY GUIDE EE 404-4, "ENVIRONMENTAL OVALIFICATION OF CONNECTION ASSEMBLIES FOR NUCLEAR POWER PLANTS" During our 324th meeting, April 9-11, 1987, the members of the Advicory Comittee on Reactor Safeguards discussed a report from our Subcomittee on Regulatory Activities regarding the proposed Regulatory Guide EE 404-4, "Environmental Qualification of Connection Assemblies for Nuclear Power Plents."

As a result of this discussion, we concur in the NRC Staff's proposal to issue the subject Guide for public coment.

After the public coment period, we expect to review the proposed final version of this Guide together with the public coments and the NRC Staff's response to them.

Sincerely, William Kerr Chainnan

Reference:

Proposed Regulatory Guide EE 404-4, Draf t 2 "Environmental Qualifica-l tion of Connection Assemblies for Nuclear Power Plants," dated September 8,1986, transmitted to the ACRS by a memorandum from Guy A.

Arlott.o to Raymond F. Fraley, dated March 24, 1987.

cc:

S. J. Chilk, SECY T. Rehm, EDO E. Beckjord, RES G. A. Arlotto, RES S

K. Aggarwal, RES C. Bartlett, RES R. Hernan, NRR 81

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%4.... +3 June 9, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS ACTION ON PROPOSEJ REVISION.2 0F REGULATORY GUIDE 1.100, "SEISMIC QUALIFICATION OF ELECTRIC AND MECHANICAL EQUIPMENT FOR NUCLEAR POWER PLANTS" During our 326th meeting, June 4-6, 1987, the members of the Advisory Comit-tee on Reactor Safeguards discussed a report frun our Subcomittee on Re-liability Assurance regarding proposed Revision 2 of Regulatory Guide 1.100, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants."

As a result of this discussion, we concur in the NRC Staff's proposal to issue the subject Guide for public coment.

After the public coment period, we expect to review the proposed final version of this Guide together with the public coments. and the NPC Staff's response to them.

Sincerely, William Kerr Chairman

Reference:

hemorandum from Guy A. Arlotto, Office of Nuclear Regulatory Research to Raymond F. Fraley, ACRS, dated May 8, 1987, transmitting:

1.

Proposed Revision 2 to Regulatory Guide 1.100, Draft 4, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power j

Plants," dated December 4,1986, and l

2.

Proposed IEEE Standard 344-1987, "Recomended Practices for Seismic Qualification of Class IE Equipment for Nuclear Power Generating Stations," Draft 9, dated July 1986 cc:

S. J. Chilk, SECY T. Rehm. EDO E. Beckjord, RES G. A. Arlotto, RES S. K. Aggarwal, RES C. Bartlett, RES R. Hernan, NRR 83

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o UNIT ED STATES NUCLEAR REGULATORY COMMISSION o

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,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

g W ASHING TON, D. C. 2055 5 August 11, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations V. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS ACTION ON PROPOSED REGULATORY GUIDE EE 006-5, "QUALIFICATION OF SAFETY-RELATED LEAD STORAGE BATTERIES FOR NUCLEAR POWER PLANTS" During the 328th meeting of the ACRS, August 6-8, 1987, we discussed a report from our Subcommittee on Reliability Assurance regarding proposed Regulatory Guide EE 006-5, "Qualification of Safety-Related Lead Storage Batteries for Nuclear Power Plants," Draft #2.

As a result of this discussion, we concur in the NRC Staff's proposal to issue the subject Guide for public coment.

After the public comment period, we expect to review the proposed final version of this Guide together with the public comments and the NRC Staff's response to them.

Sincerely, w

William Kerr Chairman

References:

Memorandum from Guy A. Arlotto, Office of Nuclear Regulatory Research to Raymond F. Fraley, ACRS, dated June 12, 1987, transmitting:

1.

Proposed Regulatory Guide EE 006-5, "Qualification of Safety-Related Lead Storage Batteries For Nuclear Power Plants,"

Draft #2 2.

IEEE Standard 535-1986, "IEEE Standard for Qualification of Class 1E Lead Storage Batteries for Nuclear Power Plants," June 25, 1986 cc:

S. J. Chilk, SECY T. Rehm, ED0 E. Beckjord, RES G. A. Arlecto, RES S. K. Aagarwal, RES C. Bart.lett, RES R. Hernan, NRR 85

l UNITED STATES

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NUCLEAR REGULATORY COMMISSION l

n i

,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

W ASHINGTON, D. C. 20555 October 14, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON PROPOSED FINAL REGULATORY GUIDE (TASK EE 404-4), "ENVIRONMENTAL QUALIFICATION OF CONNECTION ASSEMBLIES FOR NUCLEAR POWER PLANTS" During the 330th meeting of the Advisory Comittee on Reactor Safeguards, October 8-10, 1987, we concurred in the regulatory position proposed in Regulatory Guide (Task EE 404-4), "Environmental Qualifica-tion of Connection Assemblies for Nuclear Power Plants."

Mr. C. J. Wylie did not participate in the Comittee's deliberations regarding this matter.

Sincerely, William Kerr Chainnan l

l

Reference:

Proposed Final Regulatory Guide (Task EE 404-4) (May 1987 version) transmitted by memorandum dated September 21, 1987 from G. A. Arlotto, Office of Nuclear Regulatory Research, to R. F. Fraley, ACRS,

Subject:

Regulatory Guide EE 404-4, "Environ!nental Qualification of Connection Assemblies for Nuclear Power Plants."

cc:

S. J. Chilk, SECY T. A. Rehm, EDO E. Beckjord, RES G. A. Arlotto, RES S. K. Aggarwal, RES C. Bartlett, RES R. Hernan, NRR 87

1. pa arc jo UNITED STAVES ey g

NUCLEAR REGULATORY COMMISSION o

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 0,

WASHINGTON, D. C. 20555

• • ~,4 l

l May 13, 1987 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON AN IMPLEMENTATION PLAN FOR THE SAFETY GOAL POLICY During the 325th meeting of the Advisory Comittee on Reactor Safe-guards on May 7-9, 1987, we formulated comments on the NRC Staff's 5

proposed implementation plan for the Comission's August 4, 1986 Policy Statement on Safety Goals for the Operations of Nuclear Power Plants. This topic had previously been considered during our 321st, 322nd, 323rd, and 324th meetings in January, February, March, and April of 1987, and during a subcomittee meeting on January 7,1987.

In our review we had the benefit of discussions with the NRC Staff and the documents listed.

We do not consider the current Staff proposal (Reference 1) suitable as a plan for implementing the Safety Goal Policy.

Instead, we propose a plan of three elements:

1.

Use of safety goal criteria by the NRC Staff to judge the adequacy of regulation rather than to make regulatory judgments about specific plants.

2.

Recognition and formulation of an explicit hierarchical struc-ture among the interrelated criteria in the overall goal.

1 3.

Continuation of a program to make risk estimates for specific plants, as a sampling process to assist in the evaluation of regulation.

USE OF THE SAFETY G0AL It appears that the plan proposed by the NRC Staff is intended for their use in juJging whether a specific nuclear power plant can be permitted to operate or continue to operate.

If the Staff concludes that a plant does not meet certain quantitatively stated elements of 89

m l

1 The Honorable Lando W. Zech, Jr. May 13, 1987 the Safety Goal Policy Statement, an evaluation would be made of possible changes, fo-example in design, equipment, or procedures, to determine if sucn changes would result in an finproved or acceptable level of risk in a cost-effective manner.

We do not believe that probabilistic risk assessment (PRA), on which the proposed process is based, is sufficiently developed to be used to make narrowly differentiated decisions about specific plants.

Rather, the Safety Goals should be used in a more global manner to judge the suitability of existing regulations and regulatory practices, or to assist in formulating whatever changes are nacessary to provide confidence that nuclear plants are operating within an envelope established by the Safety Goals.

We continue to believe that systematic examinations of individual power plants as described in the Severe Accident Policy Statement, based in part on insights gained from risk analyses, can serve many useful purposes.

For example, the search for risk outilers for individual plants should be perforried.

We believe that detailed qualitative information on plant characteristics and behavior is an important result of such a search, but that quantitative information (such as core melt frequency estimates for an individual plant) developed by a PRA is less robust.

We are convinced also that direct participation by managerial, engineering, and operational personnel in a systematic examination of their plant can provide them with valuable insights and understanding of plant behavior in abnomal situations.

The Safety Goal Policy should be used by the NRC Staff chiefly as a standard for judging the adequacy and appropriateness of regulations and regulatory practices to assist them in reaching decisions about reaulatory requirements.

However, to make the Safety Goal Policy usable for these purposes, develorment in two treas is needed.

The multiple goals and criteria should be related more logically in a hierarchical structure, and the "sanpling" of existing plants should be expanded beyond the work which served as a basis for NUREG-1150, "Reactor Risk Reference Document," Draft for Coment, dated February 1987.

HIERARCHICAL STRUCTURE OF THE SAFETY G0AL AND IMPLEMENTATION PLAN Several goals and guidelines are included in the Safety Goal Policy.

We believe that it would be useful to present these in a hierarchi-cal structure to facilitate implementation of the Policy in a range of circumstances.

The highest level would serve as the Commission s statement of intent in regulating nuclear power and could then be used in decisions about broad policy matters and general comparisons 90

{

The Honorable Lando W. Zech, Jr. May 13, 1987 with other industrial and technological tetivities.

The lower l

levels could then be used, as development is completed, by the NRC Staff in making specific regulatory decisions.

Each subordinate level of the hierarchy should be consistent with the level above, but should be a

more practical surrogate, representing a

simplification or quantification of the previous level.

Each surrogate should not be so conservative that it creates a de facto new policy.

It should also provide a basis by Aich to assure that the Safety Goal Policy objectives are being met.

A recommended hierarchical arrangement of the multiple goals in the Policy Statement is presented below.

• Level One:

This would be the pair of qualitative goals as i

stated in the Connission Policy Statement of August 4,1986.

• Level Two:

This would be the pair of quantitative health oEjectives as stated in the same Policy Statement.

• Level Three:

This would be the previously proposed general performance guidelint that the likelihood of a large accidental release should be less than 10E-6 per reactor year.

If this general performance guideline is to serve as a surro-gate for the Safety Goal Policy objectives, as proposed in our letter of April 15, 1986, it should represent a level of safety consistent with the Level One and Two goals. A definition of a large release as one that will lead to whole body doses of 5 or 25 rem to an individual at a plant boundary, as has been given some public mention, does not satisfy this criterion.

Such a definition is so much more restrictive than the Level One and Two goals that it, in effect, establishes an alternative policy rather than serving as a more easily applied surrogate.

We believe that this is a distortion of the intent of the Policy and suggest that a consistent definition of a large release would be one that, if it occured, would result in significantly larger whole body doses.

• Level Four:

This level of the hierarchy would consis* af three

{

performance objectives to be relied on in ensuring that the safety of operating plants is consistent with the level One, Two, and Three criteria.

These objectives should be explicit enough that they could be used by the NRC Staff in making decisions about specific regulations and regulatory practices.

Such objectives are described below.

91 I

_____---____-----___-_--_____---_----_--_-a

The Honorable Lando W. Zech, Jr. May 13, 1987 (1) The first performance objective would be an expressic the effectiveness of plant accident prevention systu.ns.

We have previously recommended a goal of "?ess than 10E-4 per reactor-year" for the mean core melt f requency "for all but a few existing plants."

By core melt, we mean loss of assured core,:ooling which can result in severe core damage.

There is an unquantified, but probably substantial, difference between the probability of loss of assured core cooling and the probability of the "core on the floor stage".

The latter is more surely threatening to the health and safety of the public, bat is also less likely.

In relating this performance objective to the risk-baseo Safety Goals, one ' ill have to confront the difficult technical issues associated with the progress of a severe accident.

This will not be easy, but a core melt prob-ability objective is less useful at this level of the hierarchy if its reution to the ultimate objective is unclear.

Core melt, as defined here, is an identifiable waypoint in the development of a severe accident.

(2) The second performance objective would b2 an expression of the effectiveness o' the design of plant accident mitica-tion systems.

Between core melt, as defined above, and challenge to containment, as no:nally understood, there are several stages at which the accident sequence mgy be arrested.

A conta< ment performance objective cannot be stated simply in terms of the Level Three probability of a large release and the probability of a : ore melt as discussed above.

We recommend that as a minimum the centainment performance objective should be such that there is less than one chance in ten for a large release for the entire fs':11y of core helt scenarios.

(3) The third performance objective would be an expression of how well the plant is operated.

This remains to be developed. A separate objective of this sort would not be necessarf if operating performance were appropriately considered in the first two performance objectives.

Nr'e ve r, present methods of analysis for performance

%,ectives are based primarily on system design only.

For i'is reason it seems necessary at this time to consider 1,2 rations in a separate objective, if the Safety Goal 92

~

i The Honorable Lando W. Zech, Jr. May 13, 1987 Policy is to be applied to plant operation and not just to plant design.

We recognize this to be a major undertak-ing, but regard it as essential to a meaningful imple-mentatien of the Safety Goal Policy.

• Level Five: The final level of the Safety Goal Policy logic is the existing body of regulations and regulatory practices.

Implementation of the Safety Goal Policy, as we propose, can be viewed as a review of "Level Five" to ensure that it is con-sistent with and carries out the intent of the goal levels above it.

The overall policy implementation that we propose consists, in effect, of transforming a bottom-up system of regulation to a top-down system as the maturing of the nuclear industry and regulation and understanding of risk' have per-mitted.

In the end, as the effectiveness of the detenninistic regulations and regulatory practices is more closely related to the Safety Goal Policy, it will be appropriate to adjust the regulations and regulatory practices to make them consistent with the Safety Goals.

SAMPLING OF PLANTS 1

As indicated 6bove, our recorsnendation for implementation of the Safety Goal Policy is that it should be used principally as a measure of the adequacy of the regulations and regulatory practices.

Safety perfonnance at r.uclear power plants then should reflect to a substantial degree the success of these procedures.

Further in order to measure effectiveness of the reculations and regulatory practices, the product must be tested.

The essential difference between what the ACRS proposes and what the Staff has proposed in this regard is that we believe the measurement of specific plants against the safety goal should be explicitly recognized as a samp1-ing process. The goal of the prccess should be to determine why and how the regulations and regulatory practices have caused an individual plant or a clas: of plants to conform with or fall short of the goal, not to simply determine whether an individual plant or class of plants conforms witn or falls short of the goal.

The purpose cf the body of regulations and regulatory practices should be to provide a population of nuclear power plants that corfoc.ns to the Corrniasion's safety policy intent, as expressad by the Safety Goal Policy.

A Safety Goal Policy implementation plan structured as suggested above can and should be used by the NRC Staff in its evaluation of proposed changes in rt:gulation that arise from a variety of sources, such as operating experiences and resolution of generic issues.

However, we believe a more proactive program shoc:d be undertaken as 93

The Honnrable Lando W. Zech, Jr. May 13, 1987 part of the Polic) implementation.

This would be a prioritized, systematic review o' the body of regulations and regulatory prac-tices (i.e.,

Level Five of the recomended Safety Goal Policy hierarchy) for conforman;e with the overall Policy. Such a program would be, in some regards, a continuation of the work that has resulted in the draf t NURLG-1150 and in previous assessments of full-secpe PRAs for particular plants.

However, we believe a new program can be better focussed on sampling a sufficient number of plants and classes of plants with the aim of assessing the effectiveness of regulations and regulatory practices that have guid2d the design, construction, location, and operation of these plants.

LIMITAT10h5 We note that there must be recognition of important limitations in the irplementr, tion of the Safety Goal Policy. These limitations are essentially 1 hose of the PRA methodology used, and are caused by a fundamental inability to accurately predict and calculate precise values of risk.

Variability in data, uncertainty about applica-bility of data, imperfect understanding of important physical phenomena, and inevitable incompleteness in analysis all contribute to this limitation.

The NRC Staff must recognize the limitations of risk analysis and limitations in the definition of the Safety Goals themselves and must apply sufficient margins within its regulations and regulatory practices to accommodate these limitations. They have always.1ad to make such judgments and allowances.

The key point is that the NRC Staff and the industr,i will be better able to make bal6nced and consistent decisions atout regulation, design, and plant operation with guidance provided by the Safety Goals and PRA than without.

The development of a Safety Goal Policy has been a long and diffi-cult, but an irportant and pioneering, effort. We believe an imple-mentation plan along the lii;es we have proposed will ensure that the Policy is used effectively in regulation.

Additional coments by ACRS Henber David Okrent are presented below.

Sincerely, William Kerr Chainnan 94

The Honorable Lando W. Zech, Jr. May 13, 1987 Additional Coments by ACRS Member David Okrent 1.

The general plan proposed by the ACRS for impleme.1tation of the Safety Goal Policy seems attractive at first sight.

The ACRS recomends the use of safety goal criteria to judge the adequacy of regulations rather than to make regulatory judgments about specific plants.

The ACRS does not believe that PRA is sufficiently developed to be used to make narrowly differentiated decisions about specific plants.

One major problem with this approach, in my opinion, arises from the current USNRC backfitting rule, which says in 50.109 Part (3):

The Comission shall require the backfitting of a facility only when it determines, based on the analysis described in paragraph (c) of this section, that there is a substantial increase in the overall protection of the public health and safety or the comon defense and security to be derived from the backfit and that the direct and indirect costs of implementation for that facility are justified in view of this increased protection.

A host of more easily specified backfits was hurriedly required in the United States after the accident at Three Mile Island, some with inadequate evaluation.

However, the more difficult but often more significant issues have been deferred for years.

These have now met up with a backlash against backfitting where I fear the pendulum has swung too far.

France and the Federal Republic of Gennany (FRG) have caen maintained a disciplined program of backfitting, as well as promulgating safety improvements for nuclear plants to be built.

France and the FRG have utilized PRA methodology in a way which resembles the ACRS proposal.

Detenninistic require-ments were developed, frequently with the aid of insi.jhts obtained from PRA, to deal with perceived vulnerabilities in the overall safety approach.

Cost / benefit analysis was not ignored but does not appear to have had a dominant impact on the decision-making prtcess in France and the FRG.

I would have more hope for the proposed ACR5 arproach if the basic t%RC safety position was to achieve in tinaely fashion a reasonable assurance that the high level safety goals and quantitative design objectives were being approached or met, without undue emphasis on cost /ber.efit analysis and the test 95

The Honorable Lando W. Zech, Jr. May 13, 1987 for "a substantial increase in the overall protection" as has been practiced during the pist few years.

2.

I agree with the ACRS that an additional sampling of nuclear plants is needed in order to consider the adequacy of current regulations.

In my opinion, draft NUREG-1150 not only i, inadequate for this purpose, it is also misleading =With regard to the present level of safety of LWRs in the United States, and should not be used by the NRC to provide such a perspec-tive.

I have many reasons for this opinion, some of which follow:

• External events are not included in draft NUREG-1150 (Reference 3).

' Many other potentially important contributors to risk, such as design and ennstruction errors, aging and inade-quate qualification of equipment, certain aspects of human error, certain types of systems interactions, and the effect of poor management quality are also not included in draft NUREG-1150.

• Some of the plants studied in draft NUREG-1150 had previ-ously received the benefit of safety 19provements result-ing from one or two earlier PPAs on the same plant.

This is not the case for the majority of operaticaal LWRs.

• The PRAs in draft NUREG-1150 do not account adequately for the kinds of significant events which have occurred during the past two years or so at Surry, Brunswick, Trojan, Davis Besse, Ranch'o Seco, and TVA, among others.

• The PRAs in draf t NUREG-1150 report core melt frequencies much smaller than those estimated fo; many of the plants examined in connection with USI A-45.

Hence, not only is much additional sampling needed, but also some means must be developed for factoring into policy de-cisions the uncertainties and the significant gaps which exist in current PRAs, and for providing confidence that nuclear plants are operating within an envelope established by the Safety Goals and the supplementary objectives.

4 3.

In view of the uncertainties and imprecision in PRA results, I disagree with the ACRS position that, if the general perfor-mance guidelinu on large releases is to serve as a surrogate, it shtuld represent a level of safety consistent with the 1

96 1

The Honorable Lando W. Zech, Jr. May 13, 1987 Safety Goals.

First I should note I do not look upon the general performance guideline on the frequency of a large release as wholly or primarily a surrogate.

Furthermore, I prefer that one seek some level of assurance via performance guidelines that successively higher level goals or objectives will be met.

I question the suitability of the 'Jef'.nition of a large release which is currently proposed by the NRC Staff.

If a complemen-tary cumulative distribution function of one or more early fatalities has a chance of 10E-6 per reactor year, may not the chance of 100 early fatalities be uncomfortably high at sites with large nearby population densities?

Also, does such a proposed definition of a large release allow adequately for severe radioactive contamination of large land areas and for other relevant factors?

References:

1.

Memorandum dated January 2,

1987 from Victor Stello, Jr.,

Executive Director for Operations, to the Comission,

Subject:

"Safety Goal Implementation Status," with enclosures on Frame-work for Safety Goal Implementation, Implementation of Safety Goals in Decisionmaking for Changing Generic Requirements, and Central Issues Treated in the Safety Goal Implementation Framework 2.

U.S.

Nuclear Regulatory Comission Report, 10 CFR Part 50, "Safety Goals for the Operations of Nuclear Power Plants,"

Policy Statement, dated August 4,1986 3.

U.S.

Nuclear Regulatory Comission Report, "Reactor Risk Reference Document," NUREG-1150, Volumes 1 to 3, Draft for Coment, dated February 1987.

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WASHINGTON, D. C, 20655

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July 15, 1987 l

l The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON DRAFT NUREG-1150, "REACTOR RISK REFERENCE DOCPENT" During the 327th meeting of the ACRS, July 9-11, 1987, we discussed the draf t report NUREG-1150, "Reactor Risk Reference Document," which was issued for coment in February 1987.

The ACRS Subcomittee on Severe Accidents considered this report during meetings on January 29 and May 1,1986 and the ACRS Subcomittees on Severe Accidents and Probabilistic Risk Assessment continued the review on June 3 and July 8,1987.

In our review we had the benefit of discussions with the NRC Staff and its consultants from Sandia National Laboratory (SNL).

We also had the benefit of the documents referenced.

NUREG-1150 describes probabilistic risk analyses (PRAs) of several operating nuclear power plants.

Results of PRAs for two of these were previously reported in WASH-1400.

The plants analyzed had different containment types and included both PWRs and BWRs.

The analyses are said to be "risk re-baselining"; i.e., the methods used are current, the data used in the analyses include both generic and plant-specific inforrration, the computations n.ake use uf codes that have bean developed since the publication of WASH-1400, and the risk calculations make use of the so-called Source Tenn Code Package (STCP) that includes much of the infonnation developed by the NRC research program on severe acci-dents (althou published STCP)gh what was used was a slightly modified version of the C mtainment performance is treated in much more detail in NUREG-1150 than it was in WASH-1400.

In addition to calculations of risk attributed to internal. initiators, this report describes the results of studies which attempted to predict the uncertainties in the predictions of a number of relevant quantities, including core melt frequency and the probabilities of early and delayed l

fatalities.

In ' assessing public risk, the current version of NUREG-1150 is incom-I plete, since external accident initiators are not treated and, based l

99 t

The Honorable Lando W. Zech, Jr. July 15, 1987 on results from other PRAs, these may aroduce significant contributions to risk.

Work has begun on e<ternal 'nitiators, and later versions of the report will centain the results.

The report and its supporting documents are voluminous, and the amount of information reported is almost overwhelming. However, we believe the significance of the results and the anticipated use of the information in the regulatory process should be made explicit.

Among the conclusions reported, the following appear to be significant:

(1) The report concludes that, for the plants examined, the risk con-tributors are sufficiently disparate that no general conclusions can be drawn concerning the risk of plants not examined.

i (2) The calculated risk of each plant analyzed was less than the quantitative health effects objectives in the Safety Goal Policy Statement.

However, as mentioned above, the calculated risk did not include contributions from external initiators.

(3) The calculated risks for Surry, Unit 1, and Peach Bottom, Unit 2, were not markedly different from those reported in WASH-1400.

We were told that a number of risk-reducing improvements had been made for these plants since the original analysis, but that these were i

serrewhat offset by newly discovered risk contributors.

One of the original aims of the work reported in NUREG-1150 was to i

determine if an analysis of these selected plants would permit con-clusiord to be drawn concerning the risks of other operating plants not analyzed.

So far as we can determine from the report and from dis-cussions with the Staff, their conclusion is that these plants (and other plants that have been the subject of PRAs) are sufficiently different, and the risk contributors are sufficiently diverse, that little can be learned about one plant from the analysis of another plant, even when they are of the same general type.

This conclusion is both surprising and disturbing.

If correct, it raises serious doubts about the breadth of application of these efforts.

The Staff has not provided convincing reasons for this conclusion. More effort is needed to determine why this conclusion should be accepted, because such a conclusion would have far-reaching consequences for several Comission policies.

We have the following additional coments:

(1) We are skeptical of the method by which exptrt opinion was used in predicting uncertainties. Explanation of ard justification for the 1

100

1 The Honorable Lando W. Zech, Jr. July 15, 1987 method are obscure.

There is also reason to believe that the way j

in which the method is used can have a significant influence on the uncertainties predicted.

It is thus almost impossible to interpret the significance of the reported uncertainties or to subject them to peer review.

(2) Many of the codes used in the calculation are relatively new.

The validity of several of the codes is not well established. Further-more, nany of them have not been published and are not yet avail-able to people outside the national laboratories.

Serious peer review of the results reported is thus almost impossible.

(3)

It was emphasized by the Staff that a major contribution of the report was ti.e "insights" provided.

We recomend that these insights be better identified and that their significance for those who are not PRA practitioners be made more clear.

(4) Human performance contributions to risk (both positive and nega-tive) are f.ct well described by PRAs.

This report does not correct that deficiency.

(5) We were told by the Staff that, in light of insights developed during the work reported, resolutions or proposed resolutions of a number of Unresolved Safety Issues are to be revisited.

We recom-mend that, as an aid to understanding the report, these instances be identified.

We recommend also that the interaction between those responsible for the resolution of Safety Issues and those responsible for this report be improved.

One might conclude, both from the report and from comments made by the Staff, that the NRC regulations are inadequate to determine plant equipment and procedures necessar.v to protect public health and safety.

If this is the Staff's conclusicn, it is a drematic finding and should be emphasized, and the position developed more effectively than it is in the present draft.

If, however, regulations can be used as a mechanism i

to protect public health and safety, and we believe they can, we recom-mend that the Office of Nuclear Reactor Regulation begin early examina-tion of this report, both to apply its insights and to guide its further development.

Sincerely, William Kerr l

Chainnan l

l 101 i

i

_J

The Honorable Lando W. Zech, Jr. July 15, 1987

References:

~;

1.

U.S.

. Nuclear Regulatory Comission, HUREG-1150, Reactor Risk Reference Document. Volumes 1, 2 and 3. Draf t issued for comment, dated February 1987 2.

U.S. Nuclear Regulatory Commission, NUREG-75/104, "Reactor Safety Study - An Assessment of Accident Risks in U.S. Comercial Nuclear Power Plants," dated October 1975 (forrerly issued as AEC report l

WASH-1400).

3.

U.S.

Nuclear Regulatory Consnission, NUREG/CR-4587, "Source Tem Code Package:

A User's Guide (M001) " Battelle Columbus Labora-tory, dated July 1986.

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NUCLEAR REGULATORY COMMISSION n

I ADVISORY COMMITTEE ON REACTOP. SAFEGUARDS k

WASHINGTON, D. C. 20666 g***

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June 9,1987 Mr. Victor Stello, Jr.

Executive Director for Operations U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON PROPOSED REVISED STANDARD REVIEW PLAN SECTION 3.6.2, "DETEkMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING," DATED OCTOBER 2, 1986 In our letter to you dated November 12, 1986 concerning NRC Staff-proposed revisions to Standard Review Plan (SRP) Section 3.6.2, "Determination of Rupture Locetions and Dynamic Effects Associated with the Postulated Rupture of Piping," we reconsnend that existing SRP requirements for postulating break sizes and locations should be retained where they relate to establishing compartment and subcom-partment pressure buildup, particularly outside the primary con-tainment.

We centinue to put forth this recomendation but see no problem with the issuance of the revised SRP Section 3.6.2, provided the guidance stated in Mechanical Engineering Branch Technical Position 3-1, subpartB.1.c.(4)isimplemented. This guidance states:

If a structure separates a high energy line from an essential componert. that separating structure should be designed to withstand the consequences of the pipe break in the high-energy line which produces the greatest effect at the structure irre-spective of the fact that the above criteria might not require such a break location to be postulated.

[The "above criteria" are the criteria for postulating high-energy fluid s stems pipe l

rupture in areas other than containment penetrations.

l The retention of this general provision should assure adequate pro-tection of essential components against pipe whip, jet impingement, and the pressurization effects of high-energy line rupturet inside of compartments and subcompartments, even after the arbitrary intermedi-ate breaks are eliminated. We believe this adequately complies with our recommendation and therefore approve publication of the Federal Register Notice on the revised SRP 3.6.2.

Sincerely, William Kerr Chainnan 103

d UNITED STATES 8

- ' ','n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WA3HiNGTON, D, C. 20655 t,

\\..... * ',c June 9, 1987 l

l Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMENTS ON PROPOSED REVISIONS TO STANDARD REVIEW PLAN SECTIONS 6.5.2, "CONTAINMENT SPRAY AS A FISSION PRODUCT CLEANUP SYSTEM" AND 6.5.5, "SUPPRESSION POOLS AS FISSION PRODUCT CLEANUP SYSTEMS" During its 326th meeting, June 4-6, 1987, the merabers of the ACRS discussed the proposed changes to Sections 6.5.2, "Containment Spray as a Fission Product Cleanup System," Revision 2 and 6.5.5, "Suppression Pools as Fission Product Cleanup Systems," Revision 0, of the NRC Standard Review Plan.

These matters were also discussed during a meeting of our Nuclear Plant Chemistry Subcomittee on May 19, 1987.

As a result of these discussions, we endorse the general approach being proposed by the NkC Staff.

Sincerely, l

William Kerr Chaiman l

105

1. pa afog,'o UNITED STATES NUCLEAR REGULATORY COMMISSION o

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.,I ADVISORY COWAITTEE ON REACTOR SAFEGUARDS o

WA THINGT ON, 0. C. 20555 August 12, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON THE PROPOSED SECTION 3.6.3, "LEAX-BEFORE-BREAK EVALUATION PROCEDURES." 0F THE NRR STANDARD REVIEW PLAN During the 328th meeting of the ACRS, August 6-8, 1987, we discussed the proposed Standard Review Plan (SRP) Section 3.6.3, "Lea k-Before-Brea k Evaluation Procedures," which is intendt.d to provide detailed guidance on the implementation of the revised provisions of General Design Criterion 4, Environmental and Missile Design Bases (GDC-4) of Appendix A to 10 CFR Part 50.

As a result of this discussion, we concur in the NRC Staff's proposal to issue the proposed SRP section for public comment.

After the public coment period, we expect to review the proposed final version of this SRP section toge d er with the public coments and the NRC Staff's rerponse to them.

Sincerely, William Kerr Chairman

Reference:

Memorandum to Distribution from R. J. Bosnak, DD/DE/RES, Subj:

Resolution of ACRS/CRGR Coments on Final Broad Scope Amendment to GDC-4, dated July 29, 1987, with

Enclosures:

1. Memorandum for Commissioners from Victor Stello, Jc. EDO, Subj:

Final Broad Scope Rule to Modify General Design Criterion 4 of l

Appendix A, 10 CFR Part 50 l

2. Standard Review Plan, NUREG-0800 (Fortrerly NUREG-75/087) 3.6.3, Leak-Before-Break Evaluation Procedures 1

cc:

S. Chilk, SECY R. Bosnak, RES R. Hernan, NRR j

107

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'o UNITED STATES l"

NUCLEAR REGULATORY COMMISSION i

n 5

I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

WASHINGTON, D. C. 20555 August 12, 1987 Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comissio,n Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON PROPOSED FINAL BROAD SCOPE RULE TO MODIFi

' GENERAL DESIGN CRITERION 4, ENVIRONMENTAL AND MISSILE DESIGN BASEST (GDC-4)

During the 328th meeting of the ACRS, August 6-8, 1987, we met with representatives of the NRC Staff and reviewed the final broad scope rule to modify GDC-4.

The ACRS Subcomittee on Metal Components held meetino. on this subject on February 27-28, 1986 and July 24, 1987 with representatives of the NRC Staff and the nuclear industry.

We also had the benefit of the document referenced.

We endorse the issuance of the proposed final rule modifying GDC-4 regarding consideration of the dynamic effects of postulated pipe rttptures in a nuclear power plant's design basis.

The acceptance criteria outlined by the NRC Staff appear to be conservative enough to i

ensure that the pipes in question will leak at easily detectable rates well before complete breaks occur.

l The proposed rule states, "the Comission will permit applicants and j

licensees to justify alternative environmeatal qualification require-ments case-by-case to replace those environmental qualification re-mirements which were associated with postulated pipe ruptures...." We wish to be kept informed of any relaxation of environmental quali-fication requirements outside primary containment which are based on leak-before-break consideration.

Further, we have an interest in the possibility that a licensee may be able to demonstrate that water hammer is unlikely to occur in a given high energy system outside of primary containment, and therefore leak-before-break concepts can be applied.

Should such a situation be proposed by an applicant or licensee and found acceptable to the NRC Staff, we wish to be kept informcd.

109

Honorable Lando W. Zech, Jr.

-2 August 12, 1987 Additional counts by ACRS Members David Okrent and Glenn A. Reed are -

presented below.

Sincerely, William Kerr Chairman Additional Connents by ACRS Members David Okrent and Glenn A. Reed 1

Despite the seeming arrival of cures from time to tine, we believe that the long history of stress corrosion cracking in BWPs, and the absence of an adequate history of operating experience free from intergranular stress corrosion cracking (IGSCC.

make it ~ prudent not to permit application of "leak-before-break" to BWR high energy piping at this i

time, even if stress improvement and improved water chemistry are

present, j

Reference:

Memorandum to Distribution from R. J. Bosnak DD/DE/RES, Subj.

Resolution of ACRS/CRGR Comnents on Final Broad Scope Amendment to GDC-4, dated July 29, 1987, with

Enclosures:

1. Memorandum for Connissioners from Victor Stello, Jr. EDO, Subj:

Final Broad Scope Rule to Modify General Design Criterion 4 of Appendix A, 10 CFR Part 50 i

2. NUREG-0800 (Formerly NUREG-75/087), Standard Review Plan, Section 3.6.3, "Leak-Before-Break Evaluation Procedures" l

1 1

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o UNITED STATES 4

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NUCLEAR REGULATORY COMMISSION e

I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o

wAssiwaton, o. c. rosss

• • ..a December 8, 1987 Mr. Victor Stello, Jr.

Executive Director for Operations U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON THE REVISED ASB 3-1 0F STANDARD REVIEW PLAN 3.6.1, REQUIREMENTS FOR ARBITRARILY POSTULATED JET IMPINGEMENT EFFECTS IN THE BREAK EXCLUSION ZONE In the referenced memoranoum, Eric S. Beckjord, Director, Dffice of Nuclear Regulatory Research, has requested approval of a Federal Regis-ter notice that would eliminate from the design basis the jet impinge-i ment effects associated with the arbitrary one-square-fcot break in the break exclusion (superpipe) zone of main steam and feedwater lines outside the containment.

i

. Rased on the draft Federal Register notice and discussions with the staff, we understand' that the arbitrary one-square-foot break will be retained for environmental qualification of essential eoutpment and for structural pressurizations. Moreover, we understand that the staff will continue to enforce separation and isolation of essential equipment as the pref 2rred method of providing protection, without referrin0 to postulated jet impingement effects.

Under these conditions.; we endorse.ntblicatien of the draft Federal Register notice revising ASB 3-1 of Standard Review Plan 3.6.1.

4 Sincerely, William Kerr Chairman

Reference:

l Memorandum dated October 2,1987 for Edward L. Jordan, Chairman, CRGR l

and William Kerr, Chairman, ACRS, from Eric 5. Beckjord, Director, RES,

Subject:

Request for Approval to Publish a Federal Register Notice Revising ASB 3-1 of SRP 3.6.1.

L 111 l

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o UNITED STATES

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',t NUCLEAR REGULATORY COMMISSION U

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

.Y /

wassworow, p. c. rosss

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February 11, 1987 1

The Honorable George H. W. Bush President of the Senate Washington, D.C.

20510

Dear Mr. President:

In accordv.ce with the requirements of Section 29 of the Atomic Energy Act of 1954,.as anended by Section 5 of Public Law 95-209, the Advisory Comittee on Reactor Safeguards submits herewith its comments on the Nuclear Regulatory Comission's Safety Research Program for Fiscal Year 1988.

We note that the trend of continually decreasing funding levels for the NRC Safety Research Program over the past several years has been arrest-ed and even slightly reversed. We are heartered by this development and hope that it persists.

Sincerely, William Kerr Chairman i

113

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UNITED STATES

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g NUCLEAR REGULATORY COMMISSION 2

ADVISORY COMMITTEE ON REACTOR SAFEGUARD 4 e

W ASHINGT ON, D. C. Dette February 11, 1987 I

The Honorable Jarres C. Wright, Jr.

Speaker of the United States House of Representatives Washington, D.C.

20515

Dear Mr. Speaker:

In accordance with the requirements of Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209, the Advisory Comittee on Reactor Safeguards submits herewith its coments on the Nuclear Regulatory Comission's Safety Research Frogram for Fisc.a1 Year 1988.

We note that the trend of continually decreasing funding levels for the NRC Safety Research Program over the past several years has been arrest-ed and even slightly reversed. We ate heartened by this development and hope that it persists.

Sincerely.

William Kerr Chaiman i

i i

1 114 i

1 i

i i

REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMIS$10N SAFETY RESEARCH PROGRAM FOR FISCAL YEAR 1988 9

A REPORT TO THE CONGRESS OF THE UNITED STATES OF AMERICA BY i

THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS U.S. NUCLEAR REGULATORY COMMISSION l

l l

WASHINGTON. D.C. 20S55 l

FEBRUARY 1987 115

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INTRODUCTION Our coments below relate to the proposed program-support funding of

$103.6 million for the FY 1988 NRC Safety Research Program.

These coments are limited chiefly to research that is not being proposed but should be, or in some cases, to research that is proposed but may not be needed.

1.

REACTORSYSTEMSA3Tl 1.1 Thermal-Hydraulic Transients 1.1.1 General Coments Research planned in the themal-hydraulic area is divided into two general parts.

The first is a comprehensive program to improve the understanding of thermal-hydraulic behavior in Babcock and Wilcox (B&W) reactor systems; in its fullest form, this calls for substantial indus-try support.

The second is a more general program of code development and experimental work that does not include direct industry participa-tion.

We believe that the emphasis on B&W systems is appropriate and that the industry should provide major support for this program, as has been proposed by the Office of Nuclear Regulatory Research (RES) Staff.

It is also important that the NRC maintain a viable program of themal-hydraulic research into the foreseeable future, including integral testing, separate-effects testing, and code development.

Our cocrnents on the proposed research in the themal-hydraulic area are given below.

1.1.? Integral Facilities RES has developed a comprehensive research plan to address the technical issues and regulatory needs associated with the themal-hydraulic perfonnance of B&W plants.

Central to this plan is the proposal to construct new integral test facilities. Optimum and timely results from these facilities will depend upon financial support from the industry.

We support this plan.

1.1.3 Separ_ste Effects Our. February 19, 1986 report to the Congress included the recomendation that the NRC Study the complicating effects that water hamer may have on thermal-hydraulic transients.

The NRC and the industry have now initiated a cooprative effort in this regard.

We support this effort and expect to monitor its progress.

Funding levels appear adequate at this time.

116 3

l;

, The NRC has initiated a program to develop an infomation base for the complex themal-hydraulic phenomena involved with the "bleed and feed" process used either to remove core decay heat or to allow controlled depressurization of the primary coolant of a pressurized water reactor (PWR) plant.

We believe that subsequent to the development of the infomation base RES should allocate funding for any additional research found necessary in this area.

1 1.1.4 Code Development l

The NRC has developed a revised Emergency Core Cooling System (ECCS)

Rule (10 CFR Part 50, Appendix K - ECCS Evaluation Models) that allows the use of realistic (best estimate) evaluation models to calculate the effects of loss-of-coolant accidents (LOCAs).

However, the acceptabil-ity of realistic models rests on the d u elopment of satisfactory method-ology to detemine the overall uncertainty associated with these models.

A related effort which addresses code applicability and scaling studies also is neressary; such work is either ongoing or planned by RES.

We strongly support these efforts.

In particular, funding should be assured over the next few years sufficient to allow NRC to obtain the necessary test and analytical data, primarily through the International Code Assessment Program and cooperative international efforts such as the 20/3D Program.

1.2 Accident Evaluation The Accident Evaluation Research Program being proposed includes a significant experimental component. We believe that the relationship of this research to the severe accident regulatory issue should be made clearer than it now is.

For example, there are three major experimental programs to investigate phenomena that will be encountered (if at all) only after the reactor core has melted and has penetrated the vessel.

The experiments are related to containment heating, to core-concrete j

interaction, and to containment behavior under extreme overpressure.

)

These areas of research all bear directly on issues relating to contain-ment capability and containment failure modes.

Each of these research areas is said to be designed to reduce some of the uncertainties identified in NUREG-0956, "Reassessment of the Technical Bases for Estimating Source Tems." The magnitude of these uncertainties is being estimated in the course of the preparation of NUREG-1150, "Reactor Risk Reference Document."

However, in order to plan properly additional research to reduce uncertainties, it would appear that the Office of Nuclear Reactor Regulation (NRR), or some group, should first consider what uncertainties can be tolerated in connection with its regulatory responsibilities. We have not seen evidence of this consideration. We, and we suppose RES as well, must, under the circumstances, try to judge the relevance of the proposed research with insufficient infomation.

117 l

1

- _ _-- __ With this caveat we make the comments noted below.

These comments do i

not represent a complete coverage of the Accident Evaluation Research Program.

We use them as an example of our conclusion that more con-sideration needs to be given to what has been learned from the research of the past five years or so, and what uncertainty can be tolerated by the regulators; or, put enother way, wt'at are the questions that regu-lators have encountered or are likely to encounter in dealing with severe accidents that cannot be answered with existing infonnation.

j With diminishing resources, it is increasingly important that the research be specifically designed to address safety concerns.

• The research on containments under extreme overpressure seems well designed and should produce results that are relevant, and that will contribute to the calibration of codes being developed for a description of containment behavior.

• The work on core-concrete interaction is probably needed, but a more detailed examination of the ways in which this may affect containment failure, as well as the uncertainties attributable to incomplete understandirig of the complex phenomena that characterize this interaction, would make it more likely that the research to be l

done would answer questions that will be encountered by NRR.

• Some risk analyses conclude that high-pressure-core-melt requences may produce enough direct heating of the containment atmosphere to cause early containment failure by overpressure.

The proposed research on direct containment heating may be of considerable interest in understanding the interaction of small particles of molten metal with containment atmospheres. However, it is relevant to reactor safety only if sequences which could produce such small particles in PWRs have a sufficiently high probability.

It is our opinion that equal effort should be devoted to establishing the likelihood and the effect of direct heating events, since absent such work, the program may well be misdirected.

We also reconnend that the experimental investigation give first consideration to the possibility of atwizing the required amount of material as well as the effects of containment geometry, and to the presence of water in the subvessel cavity.

l 1.3 Risk and Reliability The funding for the Risk and Reliability Research Program continues to be directed away from the development of risk assessment methodology and toward applications.

The work in the applications area for the most part consists of sof tware development and plant-specific risk assess-ment.

However we see some danger that with the current budget con-s straints, work on applications in support of licensing efforts will be 118

,y.

4 undertaken at the expense of developmental work.

It is important that the developmental work continue.

We recoerend that the applications work be coordinated with the eventual user by a systema tic process that would encourage the user to be involved in the development of the work product. RES has taken steps to involve the industry and individual licensees in this work. We consider it appropriate that some of this applications effort should be done by the industry or individual licensees rather than by the NRC.

In our previous reports to the Congress and to the Comission, we recorrended that:

• The completeness of the current family of plant-specific probabi.

listic risk assessments (PRAs) should be examined by continued search for possible weaknesses in current probabilistic analyses, e.g., accident paths either not currently evaluated or dismissed as insignificant, which may, on closer scrutiny, prove to be very important to risk.

• An improved evaluation of the entire family of containment designs should be performed.

• Improved methods for factoring uncertainty into decision making should be developed.

l The NPC Staff is currently developing rnethods for implementing the Severe Accident Policy and the Cafety Goal Dolicy.

We believe that difficulties are being encountered in part because of the lack of l

important answers in areas such as those listed above, as well as in the treatment of external events, environmental offects, aging, management quality, and human errors.

We recomend that RES initiate additional work in such areas by reprograming funds in the FY 1987 budget and continue to support this work in FY 1988.

If necessary, this should be accomplished by deferring the applications work or finding other sources of support for this work.

l 1.4 Human Factors l

l In its recent report to the NRC, "Revitalizing Nuclear Safety Research,"

the National Research Council's Comittee on Nuclear Safety Research has joined the ACRS in criticizing the NRC for not perfonning any research in the human factors area.

A specific study to determine the need for and nature of human factors research is being perforwed by the National Research Council's Committee on Hunan Factors. The final report of this Comittee is to be provided to the NRC during July 1987. We anticipate that this report will include recontrendation for a considerable program of research effort in the human factors area.

Plans for beginning such a program should be factored into the budget by RES.

119 2.

ENGINEERING SAFETY I

2.1 Eiectrical Equipment Qualification No funding for electrical equipment qualification research has been proposed for FY 1988.

In our February 19, 1986 report to the Congress, we recorrended that this research be continued; contrary to this recom-mendatinn, the research was teminated by the NRC at the end of FY 1986.

The objective of this research is to assess the probability of survival and perfomance of aged electrical equipment when subjected to hostile environmental conditions during and following incidents, including severe accidents, fires, hydrogen burns, seismic events, and credible combinations.

The results obtained from this research are important to prevent accidents as well as to mitigate the consequences of accidents, should they occur.

We consider the continuation of this work to be vital to the nuclear safety program and again recomend that itbe reinstituted and adequately funded.

Four unique test facilities, with a combined cost of over $2 million, were constructed at Sandia especially for this program. To preserve and raintain the existing test facilities and staff experience sufficient to continue and complete the electrical equipment qualification work efficiently, it will be necessary to continue funding this work in TV 1987.

Since its inception (about 1976), more than $10 million has been spent on this program.

Furds needed to continue and complete the program are approximately $1.5 nillion in FY 1987 and $0.9 million in FY 1988.

2.2 Effects of Earthquakes on Operating plants in our previous reports, we urged that RES establish an integrated progran in this arta and coordinate closely this work with NRR and angoing industry work.

We believe that this has been accomplished effectively in the Seismic Safety Research Program. We believe that the program is well managed and will, in the near future, produce answers that will help to resolve important issues.

2.3 Primary System Integrity Cast stainless steel components in the primary system lose ductility with time in service. The implications of this phenomenon for long-ters primary system integrity are significant.

Appropriate emphasis should be placed on ascertaining the likelihood of flaws resulting from fab-rication or from service, and on developing means of assessing con.

ditions under which they could pose significant risk.

2 120 1

, 3.

WASTE MANAGEMENT 3.1 High-level Waste Although we are satisfied with most aspects of the research program on hich-level wastes, we believe that more attention needs to be directed to related work under way in other countries.

This extends beyond the formal agreements for cooperative research that the NRC Staff has developed with selected foreign groups.

The Staff should also keep abreast of activities of the Nuclear Energy Agency, Organization for Econor.ic Cooperation and Development, relative to the development of guides for demonstrating, biath directly and indirectly, the capabilities of a repository for assuring the safe retention of radioactive wastes.

1 3.2 Low-Level Waste We believe that more effort should be directed to studies that will assist the states in ranking and selecting the most appropriate disposal systems for low-level wastes, based on the nature and characteristics of the sites available and associated technological and economic consid-erations.

l 121

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April 13, 1987 The Honorable Lando W. Zech, Jr.

Chaiman U.S. Nuclear Regulatory Comission Washington, D.C.

20555 Dear Chaiman Zech.

SUBJECT:

ACRS REPORTS TO THE NRC ON THE RESEARCH PROGRAM In our February 19, 1986 report to the Congress on the NRC safety research program, we requested reconsideration of the statutory requirement for an annual report.

We now wish to propose for your concurrence a change in the nature and tining of the advice we provide to the Comission on the research program.

As required by Section 29 of the Atomic Energy Act of 1954, as amended by Section 5 of Public Law 95-209, we have submitted reports to the Congress since 1977.

The first was submitted in December 1977, but subsequent reports have been submitted in February of each year after the budget has been received by the Congress.

In 1979, the Comission requested a similar report, to be submitted each year at the time the RES budget was being prepared and considered by the Comission.

The scope and content of these reports have changed markedly over the years.

Prior to 1966, our report to the Congress was a NUREG docu-ment of 40 to 60 pages containing relatively detailed coments on many aspects of the research program, coments that were directed more to the NRC Staff than to the Congress.

Our last two reports to the Congress have been brief 5 or 6 page letter reports, with corre-spondingly less detail and based to a considerable extent on our earlier report to the Comission on the same budget.

Our reports to l

the Comission also were reduced in scope, beginning in 1982, follow-

)

ing an exchange of correspondence between then Chaiman Palladino and i

ACRS chaimen (Attachments 1-4).

1 Some recent actions or developments that suggest a need to reexanine cur role in relation to the NRC research program are discussed below.

Your letter of September 18, 1986 to D. A. Ward suggested that it would be useful for the comittee to "Advise the Comission en the effectiveness and correctness of direction of LRC's research program i

123

l l

The Honorable Lando W. Zech, Jr. April 13, 1987 i

to ensure that research is relevant to the agency's safety missiod."

We do not believe that our present annual reports are responsive to this need, even in their current abridged form. Nor is it clear that you expect such advice on an annual basis, and if so, on the current schedule based on the budget process.

In its December 8, 1986 report to the Comission entitled, "Revi-talizing Nuclear Safety Research," the National Research Council Cemittee on Safety Research recomended that the NRC shon1d erpanel an independent advisory group "... charged with independently review-ing for the director of research, from the perspective of the general principles cited in this report, the overail structure and thrust of the research program."

In SECY-87-52, "Independent Advisory Panel for the Office of Research," the EDO has recorrended the creation of a Standing Board of the National Research Council to perfonn that review.

There has been a significant reduction in resources available to the ACRS, and a further reduction is pr# posed.

Even in the abridged form of the last few years, each of the two annual reports (one for the Congress and one for the Commission) requires two freetings of the rather large Safety Research Program Subcomittee, several hours of full corm 11ttee time, and substantiel amounts of review by ACRS technical or generic subcomittees to obtain infomation and develop positions on specific portier.s of the research program.

In view of the developments mentioned above, we believe that we can best serve the Comission by reporting to you on the effectiveness and correctness of direction of those elements of the safety research program that appear to either you or us to warrant attention at any given tine.

In addition, we would think it appropriate, from time to time, to provide some perspective, not tied to specific issues, on the overall thrust and relevance of the pror; ram. These reports would not be submitted on a schedule related to the budgett process nor even on e strictly annual basis.

They would be intended to keep you t

inforned of our views on the program and to provide you with a basis for formulating and defending the research program and budget.

We will be happy to discuss this with you and the other Comission-

ers, l

Sincerely, William Kerr l

Chairman 124

l l

The Honorable Lando W. Zech, Jr. April 13, 1987 Attachments:

1.

Letter from J.

Carson Mark, Chai nnan, Advisory Comittee on 1

Reactor Safeguards, to Nunzio J.

Palladino, Chai man,

U.S.

Nuclear Reculatory Comission,

Subject:

ACRS Review and Reports on Safety Research Programs, dated October 20, 1981.

2.

Letter from Nunzio J. Palladiro, Chaiman, U.S. Nuclear Regu-latory Comission, to J. Carson Mark, Chaiman, Advisory Comit-tee on Reactor Safeguards, dated December 10, 1981.

3.

Letter from J. Carson Mark, Chai ma n, Advisory Comittee on Peactor Safeguards, to Nunzio J.

Palladino, Chaiman, U.S.

Nuclear Regulatory Comission,

Subject:

ACRS Review and Reports on NRC Safety Research Programs, dated December 14, 1981.

4.

Letter from P. Shewmon, Chaiman, Advisory Comittee on Reactor Safeguards, to Nunzio J.

Palladino, Chai man, U.S.

Nuclear Regulatory Commission,

Subject:

Procedures for ACRS Review of the NRC Long-Range Research Plan, dated June 7, 1982.

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October 20, 1981 Honorable Nunzio J. Palladino l

Chaiman U. S. Nuclear Regulatory Commission Washington, DC 20555

$UBJECT:

ACRS REVIEW AND REPORTS ON SAFETY RESEARCH PROGRAMS

Dear Dr. Palladino:

Since 1977, the ACRS has been required by the Congress to report to it an-nually on the NRC Safety Research Program. This report is prepared each year, af ter OKB has transmitted the budget request to the Congress in November, and is submitted in February before the appropriate Congres-sional comittees complete their recemnendations on the authorization bill.

Since i979, we have provided a report to the Comission on the re,earch pro-gram and its budget, usually just before the EDO budget goes to the Comis-sien for final action in July. This report has been similar in scope to the Report to Congress, although the original request from the Commission was for comments on the budget rather than a complete review of the safety research program.

In 1991, we prepared a report to the Comission on the draft Long Range Re-search Plan (LRRP). This report was in the fom of a letter rather than the fomat of the other two reports noted above. This report, too, was re-quested by the Comission, and existing procedures call for similar reviews and reports on the yearly updates of the LRRP.

We believe that our reviews of the safety research program in general, and of individual areas and projects, have been useful to both us and the RES 5taff. We believe that the Staff has been responsive in large part to our comments and recomendations.

However, we do not believe that the benefits from our reviews and reports justify the expenditure of resources by the ACRS, its Staff and consult-l ants, and by the RES Staff, that has been required to make three separate reviews each year and prepare three separate reports. We unaerstand that I

Mr. Minogue agrees with this evaluation.

ATTACHMENT 1 126

1 l

l 1

1 Honorable Nunzio J. Palladino October 20, 1981 We propose to ameliorate this situation, without reducing the extent or ef-fectiveness.of our review of the program and our interaction with the RES 1

Staff, by the following procedures:

Report to Congress. We will continue to prepare this report, as before.

It will be relatively long and relatively comprehensive, and will provide comments on the nature, scope and effectiveness of the program as well as on needs and proposed funding levels.

This report will continue to be avail-able in February, and thus can be used by the RES Staff as a basis for its update of the LRRP and its preparation of the next budget cycle.

Report to the Commission.

If requested, we will, of course, provide coments or advice to the Commission on the RES budget request or on specific portions of the safety research program or on funding levels in detail or in general.

However, we prefer not to provide evaluations and comments of the kind and scope already included in the Report to Congress. Such a report to the Commis-sion would be brief and in letter fom.

Lone Range Research Plan. The first LRRP developed was little more than a five-year projection of current programs and current needs, ar.d provided little to review in addition to the reviews we had already made of ongoing programs and those planned for the next one or two years. We believe, there-fore, that reviewing the LRRP would not be an effective use of our time un-less a more meaningful plan is developed.

We would be pleased to have your comments on these proposed changes in pro-cedures, and we will be willing to discuss them with you and the Commissioners at your convenience.

Sincerely, J. Carson Mark Chaiman 127 l

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December 10, 1981 CH AIRM AN Dr. J. Carson Mark Chaiman Advisory Comittee on Reactor Safeguards U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Dea r Dr. Ma rk :

Your letter of October 20 on the ACRS's several annual reviews of the RES progra, highlights a concern ovei the amount of time and ef fort that both the ACRS and the RES staff expend on the reviews.

RES has estimated that it expends on the order of 150 to 300 man weeks per year preparing for, participating in, and doing follow-up analysis for the ACRS reviews of its program and budget.

Your letter clearly expresses concern for the amount of ACRS time also spent in these reviews. We share with you a desire to significantly reduce the time spent on these reviews, while at the same tire not reducing the benefit that we and the Con ACRS input and guidance to our research program. gress receive from the We agree that ACRS should not have to perfom three separate reviews each year. We concur with your proposed approach to develop a plan in which the Comittee would conduct only one thorough review each year, with possibly the need for some updating by RES to keep you abreast of im-portant changes. We agree with your reco mendations regarding the report to the Commission on the RES budget recuest and the preparation of a com-prehensive report to the Congress in February of each year. However, we believe that in view of the timing of the annual report to Congress on the Research program, it would also benefit the Comission, Congress and the RES staf f if this review included consideration of the Long Range Research Plan (LRRP).

It is our intention that the annual development and refinement of the LRRP constitute the foundation for the planning of our research program.

Preparation of the LRRP pemits us to lay out directions of the research program for the coming years and to obtain user office endorsement of these program directions. At this time, the research programs are in the femative stage and are more amenable to guidance and advice than at any other stage.

The planning effort can ber.efit from the perspective of a group of experts in nuclear safety prob 1 cms who are in intimate contact with the current regulatory challenges. A thorough review by ACRS at this stage should provide all of the background and material needed to ATTACHMENT 2 128

2 December 10,1981 allow the fulfillment of your obligations to the Congress and would be su'ficient to provide ey fellow Comissioners and me the benefit of your advice for cur review of the RES budjet in the sumer. Nomally, the plan would be available for your review in December with sufficient time for you to hold succo rittee meetings in January.

The report for the l

Congress could also include an update of any thanges (mostly deletions) l to the progra-for the budget year being considered by the Congress that l

may have occurred as a result of the NRC-internal and OMB budget review process.

The LRRF for 1953-1957 (NUREG-0740) was the first attempt at preparing a plan under the current criteria.

(Five-year plans were prepared in 1976 and 1977.) RES nas received constructive criticism on that plan; the nest version, which is now being prepared, will have the benefit of that advice.

We expect that it will include more detailed program descriptions, discussion of reed end expectee use of results.

Thus, we concur with the ACRS reco rendations contained in your letter of October 20, 1951 with the exception that an ACRS review of the LRRP be in-cluded in the corerehensive review of the research program which foms the basis for ye,r ann al report to Congress. This would give us the benefit of your advice at the earliest and most productive stage and, we believe, result in the ncst efficient use of your and our RES staff time.

Com-issioner Ahearre agrees to the ACRS reducing their level of budget re-view, bst wcsid have preferred to retain some level of ACRS review and co rent on the more significant items. The report to Congress is on the budget sub.itted to Congress. This review does not duplicate the report to the Co rission on the budget being considered for future submittal.

However, give9 the nature of the Long Range Research Plan, he agrees to relieving ACR5 of their review of the Long Range Research Plan.

Please let us knca if you have any additional thoughts on this matter.

Sincerely,

')

/W-a Nunzio J. Palladino 129

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December 14, 1991 Honorable Nunzio J. Palladino Chai rman U.S. Nuclear Regulatory Comwission Washington, D.C. 20555

SUBJECT:

ACRS RLVIEW AND REPORTS 04 NRC SAFETY RESEARCH PROG %95

Dear Dr. Palladino:

In our letter of October 20, 1981 we expressed our belief "that reviewing the LRRP would not be an effective use of our time unless a more meaningful plan is developed." Although we anticipate significant improvements in the LRRP, it is perhaps too lata to use the new LRRP as a basis for our report to Congress on the FY 1983 program since that reoort is well under way, and we have not yet receivert the new plan. Nevertheless, we intend to review the plan and, to the extent needed and practicable, provide you and the Connissioners with our coaments.

It is likely that our comments this year car, be based primarily on the reviews we have carried out in preparation for our report to Congress; extensive interaction with the RES Staff should not be necessary. Nevertheless, we will consider ways in which our review of the FY 1984 Safety Research Program can be carried out in order to provide you with timely and useful :ornments on the LRRP and, at the same time, provide us with the information and insights we need to prepare our report to the Congress.

With regard to a review and report to the Commission in July on the RES budget request, we said in our letter of October 20, 1981 that we will con-g tinue to provide comments on funding levels, in detail or in general, and on specific portions of the program, in doing se, however, we would expect to limit our interaction with the RES Staff; this would be possible if there is an easily identifiable relation between their budget request and the needs and programs described in the LRRP.

Moreover, we would not intend to elaborate on the bases for our recommendations if it is possible to relate them to comments made previously in connection with the LRRP and our report to Congress.

We will continue to make both general and specific recommendations to the Commission and to the RES Staff.

It would be helpful to us in our continuing review of the Safety Research Drogram, if RES would respond in writing to each recomendation, general or soecific, made in our report to the Congress.

in summary, we believe that procedures can be developed to provide the information requested in your letter of December 10, 1981.

Sincerely, J. Carson Mark Chairman 130 ATTACHMENT 3

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wumcTcm.n c.rosss J :ne 7,1982 The Honorable Nunzio J. Palladino Chairman i

U. S. Nuclear Regulatory Ce. mission Washington, DC 20555

Dear Dr. Palladino:

Subject :

PROCEDURES FOR /4RS REVIEW W THE NRC LONG-RANGE RESEARCH PLAN During the ACRS meeting with the Comissioners on June 4,1982, you asked that we sumarize our vien on the preparation of an annual ACRS report on the Long-Range Research Pitn (LRRP).

Tae memorandum from S. Chilk, SECY, to W. J. Dircks, EDO, (COVJA 80-13) dated April 22, 1980 indicated that we should review and comment on an updated LRRP in February of each year.

Accordingly, the first ACRS report, issued April 14, 1981, was on the draft LRRP for FY 1983-FY 19S7 (NUREG-0740).

Your letter of December 10, 1981 confireed your interest in receiving MRS coments on the LRRP, and we f'

issued our second report on April 5,1952, on the draft LRRP for FY 1984-FY 1958 (NUREG-0784).

We have found the LRRP to be useful; h:*ever, the review of the LRRP has

' involved considerable ef fert for both :s and the NRC Staff. This effort is in addition to that ir tolved in our annual reports on the NRC safety research program and budge; to the Comission and the Congress.

We propose that we discontinue our forn.11 report to the Comission on the LRRP.

However, we expect to continue to receive the LRRP, both in draft and final fom, and we exnct to utilize it in our review of and report on the NRC Safety Research Program and Ivdget for the Commission and the Congress.

We would be pleased to have your comnents on this proposed change in procedures.

Sincerely,

\\.

P. Shernon Chai rmin ATTACHMENT 4

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.,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555

' July 15, 1987 The Honorable Lando W. Zech, Jr.

Chainnan U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON THE INTEGRATED SAFETY ASSESSMENT PROGRAM During the 327th meeting of the ACRS, July 5-11, 1987, we reviewed the experience gained with the NRC Staff's Integrated Safety Assessment Program (ISAP) pilot program. This topic was discussed during a Subcom-mittee rreeting on July 7,1987.

The ISAP process had also been dis-cussed during our 279th meeting in July 1983, as reported in our letter to Chairman Palladino of July 12, 1983.

In our review, we had the benefit of discussions with the NRC Staff and industry representatives and of the documents referenced.

The ISAP is intended as a cooperative program, between the NRC Staff and a licensee, which provides for the optimized resolution of multiple safety issues.

The program would permit a licensee to develop an integrated plan for implementing plant improvements in response to outstanding safety issues.

It would be based on a comprehensive list of issues, an assessment of the plant's operating experience, and a plant-The ISAP concept grew out of the Systematic specific risk analy(sis.SEP), which was highly successful in upgrading the Evaluation Program ten oldest licensed plants to con'ormance with current regulatory requirements.

l The ISAP pilot program was carried out for Millstone Unit 1 and Haddam i

Neck by Northeast Nuclear Energy Company (NNECO) and the NRC Staff. The NRC Staff has issued a draft Integrated Safety Assessment Report (ISAR) for Millstone Unit 1 and will soon issue the draft ISAR for Haddam Neck.

The NRC Staff is preparing recomendations to the Comission for the future use of the ISAP process based on an analysis of the lessons le:rned from the pilot program.

133

1 The Honorable Lando W. Zech, Jr. July 15, 1987 Experience with the ISAP pilot program demonstrates that the process is useful and provides an excellent mechanism for the implementation of regulatory requireme:.ts.

The NRC Staff and NNECO appear to agree on the usefulness of ISAP.

We recomend that the ISAP process be extended to other plants.

The NRC Staff has not yet completed the development of the procedures by which the ISAP process would be implemented more broadly.

The methods used in the pilot program were successful. However, several points need to be considered further:

(1) the application of the process to plants that were not in the SEP (2) the coordinatien with the Individual Plant Examination program under the Severe Accident Policy implementation (3) the appropriate score for the risk analysis to be used as part of ISAP We would like to review the NRC Staff's proposal for future, use of the ISAP when it is available.

Sincerely, William Kerr Chairman

References:

1.

U.S. Nuclear Regulatory Commission, NUREG-ll84, "Integrated Safety Assessment Report " Draf t Report, April 1987 2.

Northeast Utilities, "Integrated Safety Assessment Program, Haddam Neck Plant," Volumes 1 to 3 Final Report, December 1986 3.

Northeast Utilities, "Integrated Safety Assessment Program, Mill-stone, Unit 1," Volumes 1 to 3, Final Report, July 1986 4.

Northeast Utilities Service Company, NUSCO 149, "Connecticut Yankee Probebilistic Safety Study," Volumes 1 to 4. February, 1986 5.

North 2ast Utilities Service Company, NUSCO 147, "Millstone, Unit 1 Probabilistic Safety Study." Volumes 1 to 3. July 1985 134

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March 9, 1987 Mr. Victor Stello Executive Director for Operations V. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON THE ADVANCE NOTICE OF PROPOSED RULEMAKING ON THE DEFINITION OF "HIGH-LEVEL RADI0 ACTIVE WASTE" During the 323rd meeting of the ACRS, March 5-7, 1987, a discussion was held with the NRC Staff relative to the Advanced Notice of Proposed Rulemaking on the Definition of "High-Level Radioactive Weste."

This topic was also discussed during a meeting of the ACRS Subconnittee on Waste Management on February 19 and 20, 1987.

On the basis of this review, we believe that the approach being taken by the NRC Staff is reasonable. We support the Staff's efforts to base the definition of high-level radioactive waste on the associated risks; it should not be based on the source of the waste.

Such an approach will provide for better protection of the public as well as better allocation of resources.

We plan to offer additional coments on this matter after responses from the public have been received and evaluated by the NRC Staff.

Sincerely, William Kerr Chainnan

Reference:

1.

Draft Advance Notic. Tf Proposed Rulemaking, 10 CFR Part 60 Definition of "Higt.-Level o dioactive Waste" (undated -- received l

a about February 1, 1987) 135

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W ASHINGTON, D. C. 20555 Parch 9, 1987 Mr. Victor Stello Executive Director for Operations l

U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON "STANDARD FORMAT AND CONTENT" (NUREG-1199)

AND "STANDARD REVIEW PLAN" (NUREG-1200), GUIDANCE DOCUMENTS FOR THE PREPARATION OF A LICENSE APPLICATION FOR A LOW-LEVEL WASTE DISPOSAL FACILITY During the 323rd meeting of the ACRS, March 5-7, 1987, we met with the NRC Staff to discuss "Standard Format and Content" (NUREG-1199) and "Standard Review Plan" (NUREG-1200),

guidance documents for the preparation of a license application for a low-level waste disposal facility.

These documents were also discussed during a meeting of the ACRS Subcommittee on Waste Management on February 19 and 20,1987 On the basis of this review, we offer the following coments.

In general, we conclude that both of these documents are overly detailed and stringent.

Both require applicants to submit information and to develop capabilities that do not appear to be warranted by the public health risks associated with a low-level radioactive waste disposal facility.

While too detailed in some respects, the reports are not clear enough in others (for example, in the definition of a "buffer zone").

They contain requirements that may exceed current technical capabilities j

(such as the verification of the class of a given waste sample, and a j

determination of whether it contains hazardous toxic chemicals).

They also contain discussions of certain topics (such as environmental monitoring) that are so dispersed throughout the reports that they are difficult to follow.

Compounding these problems is the fact that, while these two reports cite International Comission on Radiological Pro-I tection (ICRP) Publication 30 as the basis for associated radiation dose I

assessments, the referenced NRC regulations.10 CFR Part 61, are based on ICRP Publication 2 and the standards for radiation protection as prescribed in 10 CFR Part 20.

137 i

l

Mr. Victor Stello 2

Executiv9 Director for Operations i

i We recommend that the NRC Staff simplify and clarify these two docu-i nents.

It may be useful in this effort for the NRC Staff to review any comparable U. S. Environmental Protection Agency reports prepared for the review nf facilities fer the disposal of toxic chemical wastes.

Sincerely.

t.

i William Kerr Chainnan t

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,I ADVISORY COMM)TTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C,20655 April 14, 1987 Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555 6a r Chairman Zech:

SUBJECT:

ACRS COMMENTS ON PROPOSED NUCLEAR WASTE ADVISORY COMMITTEE During the 324th meeting of the ACRS, April 9-11, 1987, we briefly discussed the recomendations provided to the Commissioners by the Executive Director for Operations on arranging for outside advice and expertise on high level waste issues (SECY-87-91, dated April 3, 1987).

Since we have serious concerns on certain aspects of this proposal, we would appreciate Comission deferral of a decision on this matter until after we have had an opportunity to meet with the NRC Staff next month and to provide you with our coments.

Sincerely, W. Kerr Chairman i

139

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i ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8

WASHINGTON, D. C. 20555

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June 9, 1987 l

i The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

Dear Chainnan Zech:

SUBJFr.T: ACRS COMMENTS ON DISPOSAL OF MIXED WASTE During the 326th meeting of the ACRS. June 4-6, 1987, we heard a report from the Division of Low-level Waste Management and Decomissioning on its efforts to develop jointly with the U.S. Environmental Protection Agency (EPA) a definition of, and acceptable methods for regulating the disposal of, "mixed waste." This subject was also discussed by our Waste Management Subcomittee during a meeting on May 18-19, 1987.

Our impression is that NRC Staff members responsible for this effort have, in cooperation with the EPA, made significant progress in resolving the relevant issues (namely, siting guidelines, design standards, and the complexities of dual regulation).

We comend then for their efforts.

We have some concerns, however, about the interpretation of the definition of "mixed waste."

If a strict interpretation results in a large increase in the wastes classified within this category, this could have a negative impact on the disposal of wastes from many facets of the nuclear industry.

Specific questions to be addressed in resolving this issue include the procedures and schedule for licensing facilities where such wastes can be disposed, the role l

of Agreement States in such activities, and how such wastes are to be handled l

in the interim.

We concur with the NRC Staff's conclusion that substantial work is still l

required for the dual issuance of EPA permits and NRC licenses, as well as dual ir.spection and enforcement activities.

We request that the NRC Staff keep us infonned as progress is made in this area.

Sincerely, William Xerr Chaiman 141

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The Honorable Lando W. Zech, Jr.

Chainnan U. S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON QUALITY ASSURANCE PROGRAMS FOR A HIGH-LEVEL WASTE REPOSITORY During the 326th meeting of the ACRS, June 4-6, 1987, we heard a report from the Division of High-level Waste Management on its review of the quality assurance (0A) program being utilized by the U.S. Department of Energy (DOE) in the development of a high-level waste (HLW) geologic repository.

This subject was also discussed during a meeting of our Waste Management Subcomittee on May 18-19, 1987 We are pleased that the NRC Staff is generally familiar with the QA lessons learned from experience in the nuclear power plant arena as described in the report of the so-called Ford Amendment Study, NUREG-1055 (Reference 1).

However, we are concerned that some prior mistakes might be repeated in the HLW repository development process.

In particular, we believe that confirmation of quality should be carried out in discrete steps throughout the lengthy HLW repository development process (e.g., through readiness reviews), in contrast to confirmation near the end of the process only. Although they are not required to do so by regulations, we urge tnat the DOE and the NRC jointly define acceptance criteria and a schedule for conducting such readiness reviews.

Specific hold points should be jointly agreed to in the HLW repository development process.

Reviews should be conducted at these hold points, using the acceptance criteria agreed to in advance, to verify quality in the development process up to that point.

Confinnation of quality sufficient for licensing would indicate that the development process is ready to proceed toward the next quality review hold point.

Failure to confinn the quality of the development process as it proceeds could result in considerable difficulty in licensing the completed repository.

Although the NRC Staff has in the past expressed serious con;; erns about certain aspects of the DOE quality assurance program, we were pleased to hear a more favorable current report.

The NRC Staff should continue monitoring DOE's activities in this area.

Of special interest to us is the so-called Q-list (Reference 2) being developed by the NRC Staff for application to 143

The Honorable Lando Zech, Jr.

2 Juns 10, 1987 DOE's program.

We look forward to receiving this document as soon as it is available in t'inal form.

Concurrent with the above, we encourage the NRC Staff to reve forward rapidly in the development of an NRC quality assurance program for application to those portio.

• of NRC activities that pertain to its independent evaluation and review of the DOE high-level waste program.

The NRC quality (e.ssurance a

program should apply, in particular, to the NRC contractors g.,

the Federally Funded Research and Development Center) involved in this work.

We request that the NRC Staff keep us informed on the progress of both the DOE and the NRC quality assurance programs.

Sincerely, William Kerr Chairman

References:

1.

NUREG-1055, "!roproving Quality and the Assurance of Quality in the Design and Construction of Nuclear Power Plants" (A Report to Congress),

dated May 1984 (Reprinted March 1987).

2.

Draft Generic Technical Positen on Items and Activities in the HLW Geologic Repository Program Subject to 10 CFR Part 60 Quality Assurance Requirements, dated July 1986 1

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The Honorable Lando W. Zech, Jr.

l Chairman U.S. Nuclear Regulatory Connission Washington, D.C. 20555

SUBJECT:

ACRS ACTION ON THE PROPOSED FINAL RULE AMENDMENTS TO 10 CFR PART 72, "LICENSING REQUIREMENTS FOR THE INDEPENCENT STORAGE OF SPENT NUCLEAR FUEL AND HIGH-LEVEL RADI0 ACTIVE WASTE" During the 329th eeeting of the ACRS September 10-12, 1987, we dis-cussed a report from our Subcomnittee on Spent Fuel Storage regarding the proposed Final Rule Amendments to 10 CFR Part 72, "Licensing Re-quirements for the Independent Storage of Spent Nuclear Fuel and High-Level Radioactive Waste."

As a result of this discussion, we concur with the NRC Staff's proposal to issue the subject Final Rule Amendments.

However, we note that operators of equipment and controls at facilities licensed under 10 CFR Part 72 would not be licensed by the NRC but are reavired by NRC regu-lation to be certified by the licensee.

Because these facilities could contain large quantities of special nuclear inaterial, qualification and certification of operators is appropriate.

Although NRC requires certification of operators, we are not aware of any NRC guidance or criteria that would be used to determine the adequacy of the cualifica-tiens, training, continuing training, and certification of these opera-tors.

We believe that NRC should develop such guidance and criteria, and we would like to review them before the licensing of such facili-ties.

Sincerely, i

William Kerr Chairman

Reference:

Memorandum from Robert Bosnak, Deputy Director, Office of Nuclear Regulatory Research, dated July 14, 1987, with enclosed proposed Final 145 l

l J

'I The Honorable Lando W. Zech, Jr. September 17, 1987 t

Rule Amendments to 10 CFR Parts 2, 19, 20, 21, 70, 73, 75, 150 and Part 72,'"Licensing Reqairements for the Independent Storage of Spent Nuclear Fuel and High-level Radioactive Waste."

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S. J. Chilk, SECY i

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November 10, 1987 The Honorable Lando W Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C.

20555

SUBJECT:

ACRS COMMENTS ON RADI0 ACTIVE WASTE MANAGEMENT RESEARCH AND OTHER ACTIVITIES During the 331st meeting of the Advisory Comittee on Reactor Safeguards, November 5-7, 1987, we discussed several high-level (HLW) and low-level (LLW) radioactive waste management research activities.

We had previously discussed these activities with the NRC Staff during our 329th meeting, September 10-12, 1987 These matters were also discussed during meetings of the Waste Management Subcomittee on August 17-19, 1987 and October 15-16, 1987, and during the Subcomittee's field trip to the University of Arizona on July 28, 1987.

The recent changes in organization of the waste management activities of the NPC have provided opportunities for the proper focusing of attention, both by the NRC Staff and the ACRS, on the LLW and the HLW programs.

On the basis of our review of these activities, however, we have noted several potential problems that need to be addressed.

In this regard, we offer the following comments.

j One of our more important observations is that there is a N ed for the NRC Staff to better define the scientific bases for some of the requirements specified in various Technical Positions and the connection between these requirenents and the NRC regulations they are designed to support.

In some cases, these requirements appear to have been introduced only for the convenience of Agreement States or the operators of shallow land burial facilities.

We believe that this practice should be carefully examined to determine whether it establishes an undesirable precedent and whether such needs by the States could be accomodated by a method other than the exercise of regulatory power.

An example of this problem is the Technical Position on Low-level Waste Form.

This document demonstrates a need by the NRC Staff to define more clearly the connection between the requirements for testing the waste form and the regulations governing its performance.

We recomend that the Division of Low-Level Waste Management and Decomissioning (DLLWMD) Staff l

reexamine the fundamental bases that led to the fomulation of the Techni-cal Position and its requirements, and ensure that the test and performance requirements are pertinent to the conditions likely to be found in shallow land burial facilities.

For example, leach testing is now being required 147

The Honorable Lando W Zech, Jr. Novembet 10, 1987 i

of the LLW form.

The NRC Staff, however, was not able to demonstrate an explicit connection between this requirement and regulatory criteria.

The Staff should be directed either to define such a connection or to delete this requirenent.

Further, they should document and make readily available the analyses that form the bases of performance evaluation and accepttnce of the waste form.

The continued aging of U.S. nuclear power plants makec it likely that the volurnes of LLW from decontamination and decomissioning activities will increase.

We believe that the complexity of the chemistry of such wastes requires that the DLLWMD Staff formulate very clearly the associated problems and the proper approaches for solving them.

As e part of this process, the Staff should seek the support of consultants and/or members of the Office of Nuclear Regulatory Research (RES) contractor staffs.

We reviewed several RES programs dealing with the integrity of the HLW r

repository.

While the results of the NRC research may need to be used in an adjudicatory hearing involving the Department of Energy our review revealed that the NRC data were obtained under a quality assurance (OA) program considerably weaker than that imposed by NRC on DOE, We believe that the Division of High-level Waste Management should actively review the NRC research programs and their output, and implement such disciplined QA activities as are needed to provide data with credibility comparable to those of DOE.

Finally, the review of the RES programs revealed that only a very modest level of peer review had been employed.

Further, we note that the request for proposal for the Federally Funded Research and Development Center (now called the Center for Nuclear Waste Regulatory Analyses) appeared to discourage the contractor from publishing his results in refereed journals, thereby disallowing the usual form of peer review.

In addition to encouraging journal publication, we believe that the Staff should implement a careful, focused, and visible peer evaluation of both the quality of the research results and their applicability to regulatory requirements.

Such evaluations should be initiated for each program to the extent feasible.

should t,e periodic, and should be designed to provide clear objectives for the management of the research program.

Sincere'y.

William Kerr Chainnan 148

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NUREG-1125. Volume 9 l

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A Compilation o<

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It also includes a report to the Congr, s on le NRC Safety Research Progrrm for FY 1980.

All reports nave be made a ilable to the public through the NRC Public Document Room and e U.S. Li ry of Congress. The reports are divided ir,to two groups:

P 1: ACRS Re ts on Project Peviews, and Part 2: ACRS Reports on Gen c Subjects.

' art I contains ACRS reports alphabetized by project fnare and within

.oject nane by chronological order.

Part 2 catego ftes the reports by he trost appropriate generic subject area and with subject area by chrono gical order, i

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