ML20214A914

a Compilation of Reports of the Advisory Committee on Reactor SAFEGUARDS,1986

ML20214A914
Person / Time
Issue date:	04/30/1987
From:	Advisory Committee on Reactor Safeguards
To:
References
NUREG-1125, NUREG-1125-V08, NUREG-1125-V8, NUDOCS 8705200003
Download: ML20214A914 (217)

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1 NOTICE Availability of Reference Materials Cited in NRC Publications Most documents cited in N RC 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. Government Printing Office, Post Office Box 37082, Washington, DC 20013 7082 '

3, The National Technical Information Service, Springfield, VA 22101 Although the listing that follows represents the majority of documents cited in NRC publications, it is not intended to be exhaustive.

Referenced documents available for inspection and copying for a fee from the NRC Public Docu-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 Reoorts; vendor reports and correspondence; Commission papers;and applicant and licensee documents and cortcspondence.

The following documents in the NUREG series are available for purchase from the GPO Sales Program: formal NRC staff and contractor reports, NRC-sponsored conference proceedings, and NRC booklets and brochures. Also available are Regulatory Guides, NRC regulations in the Code of 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 and reports prepared by the Atomic 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, journal and periodical articles, and transactions, federal Register notices, federal and state legislation, and congressional reports can usually be obtained from these libraries.

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 copics of NRC draf t reports are available free, to the extent of supply, upon wr6tten request I to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Com.

mission, Washington, DC 20555.

Copies of industry codes and standards used in a substantive manner in the NRC regulatory process are maintained at the NRC Library, 7020 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 Standards, from the i American National Standards Institute,1430 0 roadway, New York, NY 10018.

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NUREG 1125 l z.

Volume 8 *( / l l

A Compilation of

Reports of

! The Advisory l l Committee on

! Reactor l Safeguards  ;

1986 Annual l

! U.S. Nuclear Regulatory i

Commission e

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ABSTRACT This compilation contains 58 ACRS reports submitted to the Connission or to

' the Executive Director for Operations during calendar year 1986. All reports have been made available to the public through the NRC Public Document Room and the U.S. Library of Congress. No classified or other controlled information was prepared in 1986. 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 narre and

, within project name by chronological order. Part 2 categorizes the reports J by the most appropriate generic subject area and within subject area by chronological order, i

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PREFACE The enclosed reports represent the recomendations and coments of the U.S.

Pluclear Regulatory Comission Advisory Comittee on Reactor Safeguards during calendar year 1986. This publication Volume 8, is an annual supplement to NUREG-1125. Previous issues of NUREG-1175 are as follows:

Volume inclusive Dates 1 through 6 September 1957 through December 1984 7 Calendar Year 1985 i

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l j ACRS MEMBERSHIP (1986) i Mr. David A. Ward

! CHAIRMAN:

E.I. du Pont de Nemours & Company Savannah River Laboratory r I

i VICE CHAIRMAN: Dr. Harold W. Lewis University of California, Santa Barbara MEMBERS: Dr. Max W. Carbon University of Wisconsin Mr. Jesse C. Ebersole, Retired l Tennessee Valley Authority

. Dr. William Kerr

! University of Michigan Dr. J. Carson Itark, Retired i

Los Alamos Scientific Laboratory 1

Mr. Carlyle Michelson, Retired Tennessee Valley Authority and AEOD/US Nuclear Regulatory Commission Dr. Dade W. Moeller Harvard University I

Dr. David Okrent 1 University of California, Los Angeles Mr. Glenn A. Reed, Retired Point Beach Nuclear Power Plant Wisconsin Electric Power Company I

Dr. Forrest J. Remick The Pennsylvania State University Dr. Paul G. Shewmon Ohio State University Dr. Chester P. Sicss, Retired (Prof. Emeritus)

University of Illinois i Mr. Charles J. Wylie, Retired I Duke Power Company

, MEMBER EMERITUS: Mr. Harold Etherinaton, Retired Jupiter Florida i

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l TABLE OF CONTENTS 1

VOLUME VI!!

P,agg fif ABSTRACT ..............................................................

v PREFACE ...............................................................

HEMBERSHIP ............................................................

vfi PART 1: ACRS REPORTS ON PRNECT REVIEWS Babcock & WITCox Owners Group ....................................... 1 i

3 l Clinton Nuclear Power Station .......................................

Davis-Pesse Nuclear Power Station, Unit 1 ........................... 5 j GESSAR-II BWR/6 Nuclear Island Design ............................... 7 i

Hope Creek Genera ting S ta tion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 19 McGu i re Nucica r Sta tion Uni ts 1 and 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

North Anna Power Station, Units 1 and 2 ............................. 21 t

Paluel - france /U.S. PWR Design Comparison .......................... 23 Perry Nuc1 car Power Plant, Unit 1 ................................... 25 South Texas Projec t , Uni ts 1 and ? . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 77 Three Mlle Island Station, Unit ? ................................... 29 PART 7: ACRS REPORTS ON GENERIC SUBJECTS Advanced Reactors ACRS Coments Regarding NRC Review of Advanced Reactor Ocsigns April 16, 1986 .......................................... 11 ix

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TABLEOFCONTENTS(CONT'D) 4 ERSE l 4

Class 9 Accidents ACRS Comments on the Implementation Plan for the

! Severe Accident Policy Statement and Regulatory Use of New Source-Term Information, March 18, 1986 ................... 35 ACRS Comments on NUREG-0956, " Reassessment of the Technical Rases for Estimating Source Terms" --

Review Copy, June 10, 1986 ....................................... 39 l

ACRS Comments on Proposed BWR Mark I Containment Requirements for Severe Accidents. December 17, 1986 ............. 43 l Decay Heat Removal Systems 1

See " Generic !ssues/USI" ......................................... 59 Emergency Core Cooling Systems j See " Rules and Regulations" ...................................... 175 1

Emergency Planning

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l Emergency Planning Requirements for NRC Licensees. l l July 16, 1986 .................................................... 45 j Protective Action Guides for Nuc1 car Emergency Planning, November 10, 1986 ...................................... 47 l

i fire Protection ACRS Views on Fire Protection Pescarch and Fire-Related Systems Interactions, July 16, 1986 .............................. 49 l i

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TAALEOFCONTENTS(CONT'D) ,

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t Generic issues / Unresolved Safety issues ACRS Coments on Proposed Resolution of U5! A-17, >

" Systems Interactions in Nuclear Power Plants," 51 May 13, 1986 .....................................................

ACRS Coments on the Pesolution of US! A-46,

" Seismic Qualification of Eouipment in Operating 55 Plants," September 17, 1986 ...................................... l ACRS Coments on Proposed Resolution of Generic i Issue 124, " Auxiliary Feedwater System Reliability," l 59 September 17, 1986 ...............................................

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! ACRS Coments on the Prioritization of the Fourth 61 i Group of Generic !ssues, November 13, 1986 .......................

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Human Factors ACRS Coments on the NRC Policy Statement on Fitness  ;

j j for Duty of Nuclear Power Plant Personnel, August if, ,

71 1986 .............................................................

Metal Components i

ACRS Coments on Draf t Technical Peport on Guidelines for RWR Coolant Pressure Boundary Piping, March ifl, i 73 [

1986 .............................................................

4 ACRS Coments on Salvaging of Contaminated Smelted l l Alloys, May 13, 1986 .............................................

75 l 1

ACRS Coments on Degraded Piping Research, l 77 September 17, 1986 ...............................................

l Operating Organfra_tions_

i I ACRS Report on the Tennossee Vallny Authority's Management Reorganization and Shutdown of TVA's Nuclear Power Plants, August 12, 1986 ............................ 79 i

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TABLE OF CONTENTS (CONT'D)

Page Procedures - ACRS/ Regulatory / Legal Responses to Recomendations of Panel on ACRS Effectiveness, March 20, 1986 .................................... 85 ACRS Coments on the Draft Commission Policy Statement >

on Technical Specifications, July 15, 1986 ....................... 105 ACRS Coments on Proposed NRC Standardization Policy Statement, August 12, 1986 ....................................... 107 ACRS Coments on Draf t NUREG-1225. " Implementation of NRC Policy on Nuclear Power Plant Standardization,"

October 15, 1986 ................................................. 111 ACRS Suggestions for an NRC Lon9 Range Plan, October 15, 1986 ................... .............................. 113 g L

ACRS Report on Proposed Policy Statement on Deferred Plants, December 16, 1986 ........................................ 131 Oualification Systems /Fquipment l See " Generic Issues /U$l" ......................................... 55

_ Radiological E_ffects See " Rules and Regulations" ...................................... 163 ACRS Coments on Quantification of Public Health l Risks. April 16, 1986 ............................................ 133 i Support of Radiation Protection Or May 13, 1906 . . . . . . . . . . . . . . . ................................

. . . . . .gan t ra tions , 135 Additional Recomendations on the Development of  :

de minimis Levels, July 16, 1986 ................................. 137 l ACRS Coments Regarding Susport of Padiation Protection '

Organfrations such as the lational Council on Radiation Protection and Measurements INCRP), the International Comission on Radiological Protection (ICRP), and the National Academy of Sciences (NAS), July 16, 1986 ................ 139 xil

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Regulatory Guides t

) ACRS Action on the Proposed Revision 3 to Regulatory Guide 1.63, " Electric Penetration Assembifes in Containment Structures for Nuclear Power Plants " 143 4

January 13, 1986 .................................................

ACRS Action on Regulatory Guide 1.114. Revision 2, '

" Guidance to Operators at the Controls and to Senior Operators in the Control Room of a Nuclear Power Unit "  !

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August 12, 1986 .................................................. l i l ACRS Action on the Proposed Revision 3 to Regulatory l

r.uide 1.63, " Electric Penetration Assemblies in l 2

Containment Structurns for Nuclear Power Plants." 147 i

)< December 16, 1986 ................................................ l 1

ACRS Comments on Proposed NRC Final Pepulatory Guide l Entitled, " Format and Content of Plant-Specific l 1

Pressurized Thermal Shock Safety Analysis Reports for l l Pressurized Water Reactors," dtd 6/86, December 16, 1986 ......... 149 l l

Reliability _and Prnbabilistic Analysi_s 1

ACRS Comments on Proposed Safety Goal Policy, March 19, .

151 l 1986 .............................................................

I Additional ACR$ Co m!nts on Proposed NRC Safety Goal ,

i Policy Statement April 15, 1986 ................................. 155 l Application of NRC Safety Goals in t.fcensing Is'ues,

) November 10, 1906 ................................................ 161 i

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?.aSe Rules and Regulations ACRS Coments on Proposed Revision of 10 CFR Part 20

" Standards for Protection Against Radiation,"

l February 19. 1986 ................................................ 163 ACRS Coments on Proposed Broad Scope Rule Revision to General Design Criterion 4, March 19, 1986 .................... 169 '

ACRS Action on the Proposed Revisions to Sections 9.2.1 and 9.2.2 of the Standard Review Plan (SRP),

July 15, 1986 .................................................... 173 ACRS Coments on the Proposed Revision to the ECCS Rule in 10 CFR 50.46. " Acceptance Criteria for ECCS for Light Water Nuclear Power Reactors," and Appendix K.

"CCCS Evaluation Models," September 16, 1986 ..................... 175 ACRS Coments on Proposed Revisions to Standard Review Plan Sections 6.5.2, " Containment Spray as a Fission Product Cleanup System," and 6.5.3, " Fission Product Control Systems and Structures " October 15, 1986 ................ 177 ACRS Coments on Propnsed Revised Standard Review Plan i Section 3.6.P. " Determination of Rupture locations and Dynamic Effects Associated with the Postulated Rupture of Piping." dtd 10/2/86, November 12, 1986 ....................... 179 ACRS Coments on the Interpretation of 10 CFR Part 50, General Design Criterion 4, " Environmental and Missile Design Pales," Decerher 17, 1986 ................................. 181 Safeguards __and Security (Sabotaan) '

i l ACRS Coments on Proposed Insider Safeguards Provisions, February 19, 1986 ....................................

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TARLEOFCONTENTS(CONT'D)

Page Safety Research Review and Evaluation of the Nuclear Regulatory Comission's Safety Research Program for Fiscal 185 Year 1987. february 19, 1986 .....................................

ACRS Connents on the NRC Safety Research Program and fludge t for Fiscal Year 1988, June 11, 1986 . . . . . . . . . . . . . . . . . . . 205 Waste Management ACHS Coments on the Definition of Low-Level Radioactive Waste Pay 13, 1986 .................................. 213 ACRS Coments on Various NMSS and PES Waste flanagement Topics, August 13, 1986 ............................... 215 ACRS Coments on the NRC Staff Review of DOE's Final Environmental Assessments of High-Level Waste Repository Sites, December 16, 1986 .............................. 229 i

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o UNITED STATES i

'g  !

8 o NUCLEAR REGULATORY COMMISSION

$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

! July 16,1986 Mr. Victor Stello, Jr.

- Executive Director for Operations

U.S. Nuclear Regulatory Commission Washington, DC 20555 j

SUBJECT:

ACRS COMMENTS ON THE BABC0CK & WILC0X (B&W) OWNERS GROUP i

SAFETY AND PERFORMANCE IMPROVEMENT PROGRAM During its 315th meeting, July 10-12, 1986, the Advisory Committee on Reactor Safeguards discussed the B&W Owners Group Safety and Performance

{ Improvement Program. The ACRS Subcommittee on Babcock & Wilcox Reactor Plants met in Washington, D.C. on June 25, 1986 to discuss the program.

The Subcommittee had the benefit of discussions with representatives and a consultants of the B&W Owners Group (BWOG) and the NRC Staff. We also had the benefit of the documents listed, l

Recent events at some B&W plants have resulted in increased NRC concern l

regarding the frequency of reactor trips and complexity of the tran-j sients in B&W plants. These concerns have led the NRC Commissioners and

' the Staff to call for a broad reassessment of B&W plants to assure that j they provide acceptable levels of safety. At the request of the Staff, the BWOG has taken a lead role in performing the reassessment.

At the time of our Subcomittee meeting the BW0G program's main emphasis j seemed to be directed at improving plant on-line performance, rather i than addressing the safety objectives of the NRC-B&W reassessment

initiative. Our review of this program indicates that it may lead to 1 l improved plant on-line performance; however, we are concerned that plant I safety does not appear to be its central focus. We believe it should be. While it is true that improved plant performance could represent j safer operation, that is not an inescapable outcome.

We offer the following additional observations and recommendations:

1. An examination of the operating history of B&W plants indicates that three B&W plants, operated by one utility, have operated with little cause for concern. The incidents that have produced con-I cerns have occurred at plants operated by several other utilities.

! It seems logical, in seeking root causes of substandard perfor-

mance, to look at the effect of operating organizations on system performance, rather than concentrating entirely on system design.

4 2. There is the observation, primarily from analysis but partially l

confirmed by experience, that B&W systems respond differently, perhaps less favorably, to upsets than do the pressurized water 1

i__ _ - _ _ _ ___ ___ . - -- _ _

Mr. Victor Stello, Jr. July 16, 1986 reactor (PWR) plants with Combustion Engineering or Westinghouse they were reactor systems. This should not be surprising --

designed to respond differently. The once-through steam generator, the integrated control system, and different piping arrangements and auxiliary capacities give a nuclear steam supply system that is more quickly responsive to load changes and other external chal-1enges than the other PWRs. Whether, from the perspective of safety, this is good, bad, or indifferent is not yet clear. The NRC Staff and the BWOG should focus on this observation and come to  !

an engineering determination as to its significance.

3. We are concerned that apparently little attention is being given to decay heat removal. We note that, even given a complex transient, if the ability to trip the reactor and remove decay heat is pre-served the ability to protect the public is ensured.

We expect to meet again with the BWOG and the NRC Staff to continue discussions.

Sincerely,

. k David A. Ward Chairman

References:

1. B&W Owners Group Trip Reduction and Transient Response Improvements Program, BAW-1919 Revision 00, May 1986

2. Letter from D. Crutchfield, NRC, to H. Tucker, BWOG,

Subject:

B&W Design Reassessment, dated June 2, 1986

3. Letter from H. B. Tucker, BWOG, to C. Wylie, ACRS,

Subject:

B&W Owners Group Safety and Performance Improvement Program, dated July 10, 1986 I

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'o,, UNITED STATES

! 7, NUCLEAR REGULATORY COMMISSION 3  : I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o, [ WASHINGTON, D C. 20555 f

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'% * * * * * / October 16, 1986 l

MEMORANDUM FOR: Victor Stello, Jr.

E tive Director for Opers tions l

FROM: Da'n ..W i

, ha 'f Advisory Committee on Reactor Safeguards

SUBJECT:

CLINTON NUCLEAR POWER STATION - RESOLUTION OF ACRS COMMENTS During our 318th meeting, October 9-11, 1986, we were given a presenta-tion by the NRC Staff and by representatives of the Illinois Power Company regarding items that had been mentioned in a Committee report of March 9,1982 to Chairman Palladino concerning Illinois Power Company's request for an operating license.

The resolution of those items was, in our view, appropriate. We have no further questions concerning the operating license for this station.

We appreciate this information having been provided to us.

cc:

R. Bernero, NRR W. Houston, NRR W. Butler, NRR B. Siegel, NRR 3

p sts

'o UNITED STATES g

! o NUCLEAR REGULATORY COMMISSION

$ . ,E ADVISORY COMMITTEE ON REACTOR SAFEGU ARDS 0 W ASHINGTON, D. C. 20555 July 16, 1986 l Honorabla Lando W. Zech, Jr.

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

Dear Mr. Zech:

SiiBJECT: ACRS COMMENTS ON THE RESTART OF DAVIS-BESSE NUCLEAR POWER STATION, l UNIT 1 During its 315th meeting, July 10-12, 1986, the Advisory Committee on Reactor Safeguards completed its review of the actions taken to resolve concerns raised as a result of the loss of feedwater event at the Davis-Besse Nuclear Power Station on June 9, 1985. The Committee had the benefit of discussions

, with representatives of the Toledo Edison Company (Licensee) and of the NRC Staff. A Subcommittee toured . the facility and met in Oak Harbor, Ohio on October 4, 1985. This matter was previously considered during the 306th ACRS

meeting, October 10-12, 1985, and the 309th ACRS meeting, January 9-11, 1986, I and at Subcommittee meetings on February 6 and June 27, 1986 in Washington, D.C. During our review, we had the benefit of the documents referenced.

Following the event at the Davis-Besse Plant on June 9, 1985, the Director of the Office of Nuclear Reactor Regulation issued a 10 CFR 50.54(f) letter identifying general areas of concern which must be addressed by the Licensee in its response to the NRC regarding this event, and confirming that restart of the plant was subject to prior NRC approval.

! In response, the Licensee filed the Davis-Besse Course of Action report dated i September 10, 1985, with subsequent amendments. The NRC Staff's evaluation of the Course of Action report' is contained in the Safety Evaluation Report Related to the Restart of Davis-Besse Nuclear Power Station, Unit 1, Follow-ing the Event of June 9, 1985 (NUREG-1177).

The Licensee has conducted an extensive investigation of the event and has made major modifications to its management and maintenance structures.

Further, the Licensee has undertaken an in-depth System Review and Test Program, has initiated a configuration management program, and has undertaken major plant improvement and modification programs. Completion of some of these improvements and modifications is expected to extend beyond plant res tart.

One of the proposed improvements is the addition of an emergency depres-surization system which, in view of certain characteristics of this plant, such as its not totally safety grade electric auxiliary feedwater pump, its 5

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Honorable Lando W. Zech, Jr. July 16, 1986 intermediate pressure safety injection system, and its single train PORV blowdown arrangement, appears to have merit if properly engineered. We wish to be kept informed of the NRC Staff's review.

Based on our evaluation of the Licensee's actions and our evaluation of the review of these actions by the NRC Staff, we conclude that, if the Licensee implements the comitments identified in NUREG-1177, the Davis-Besse Nuclear Power Station can resume operation without undue risk to the health and safety of the public.

Though not directly related to the restart issue, we note that we expressed concern in our letter of November 14, 1985 that the implications of the incident investigation of the Davis-Besse event seemed not to be proceeding to a focused conclusion. Chairman Palladino assured us, in his letter of January 23, 1986, that this would be accomplished in the SER for Davis-Besse.

We are considering this SER as part of our continuing interest in the Inci-dent Investigation Team process.

Sincerely, David A. Ward Chairman

References:

1. U.S. Nuclear Regulatory Comission, " Safety Evaluation Report Related to the Restart of Davis-Besse Nuclear Power Station, Unit 1, Following the Event of June 9,1985," USNRC Report NUREG-1177, dated June 1986

2. Davis-Besse Course of Action, September 10, 1985, with Amendments 1-8

3. U.S. Nuclear Regulatory Commission, " Loss of Main and Auxilitry Feed- i water Event at the Davis-Besse Plant on June 9,1985," USNRC Report )

NUREG-1154, dated July 1985

4. IE Infonnation Notice No. 85-50, " Complete Loss of Main and Auxiliary Feedwater at a PWR Designed by Babcock & Wilcox," dated July 8, 1985 l

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8 pnagg'o UNITED STATES P o NUCLEAR REGULATORY COMMISSION U

-E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e wasamcTow. o. c. 20sss January 14, 1986 Honorable Nunzio J. Palladino Chairman

U. S. Nuclear Regulatory Comission I

Washington, D. C. 20555

SUBJECT:

ACRS REPORT RELATED TO THE FINAL DESIGN APPROVAL OF THE GESSAR II BWR/6 NUCLEAR ISLAND DESIGN APPLICABLE TO FUTURE PLANTS During its 309th meeting, January 9-11, 1986, the Advisory Committee on Reactor Safeguards completed a review of the reference design described in the General Electric Standard Safety Analysis Report (GESSAR II) for a Final Design Approval (FDA). GESSAR II provides the safety information for a reference system consisting of a single BWR/6 Mark III nuclear steam supply system, with a design power level of 3730 MWt, and associated systems and structures, including the reactor building (the shield building and containment), fuel building, diesel generator buildings, control building, auxiliary building, and radwaste building.

Subcomittee meetings were held with representatives of the General Electric Company (the Applicant) and the Nuclear Regulatory Comission (NRC) Staff on October 18-19, December 4-S,1984, and February 14-15, 1985 in Los Angeles, Calif.; on March 27-29, 1985, in Albuquerque, New Mexico; and on August 7 and September 11, 1985 in Washington, D. C. The full Comittee considered this matter during its 299th through 309th meetings held monthly from March 1985 through January 1986.

We believe that the GESSAR II design includes features that provide a significant improvement in safety over current BWR designs. If this were an application for a construction permit for one or more plants of this design, we would have no hesitation in recomending its approval.

However, we are unable to agree with the Staff, for reasons discussed below, that the design satisfactorily or completely addresses all of the concerns described in the Commission's Severe Accident Policy.

While there is no doubt that, in the future, new plants should be consistent with the Severe Accident Policy, we see no harm in the approval of the GESSAR II design, provided that this approval is for a limited time (say five years), and provided that this procedure not be viewed in any way as a precedent for the handling of future applica-tions. In particular, the information provided to us in connection with GESSAR II would not be sufficient to support an application for a one-step license.

Our concerns about the review and the review process are elaborated in the following paragraphs. '

7

)

1 Honorable Nunzio J. Palladino January 14, 1986 l

1 i We believe that reviewing the GESSAR II design under the Severe Accident i Policy was premature and incomplete. We do not see how the Severe Accident Policy can be implemented for an FDA while the policy on safety l goals is still in the process of being developed. The NRC Staff's severe accident review of the GESSAR II design was based on the accept-I ance of values of core-damage probability and the use of cost / benefit

analyses that may turn out to be quite different from those adopted by I the Commission for implementation of the safety goals. These and other i

concerns are discussed in the following items:

. In its policy statement on severe accidents, the Commission did not

provide detailed guidance to the Staff concerning the safety philosophy the Commission desires for future plants. The require-ment for completion of a PRA and of a Staff conclusion of safety j acceptability leaves the matter of desired safety level undefined and something to be decided ad hoc for each future plant or stan-f dard plant design application.

. The Applicant and the Staff both evaluated the cost / benefit ratio j of a large number of potential safety improvements. However, the <

j approach used by the Staff is that which has been used in the past i and may or may not be that which the Commission will adopt in its j continuing consideration of its Safety Goal Policy.

r j . We believe that further evaluation is needed regarding the likeli-

hood of loss of containment integrity, given an accident leading to melt-through of the reactor pressure vessel. Should this likeli-hood be large, as the Staff says it is, the acceptability of such a t characteristic of containment behavior for a future plant should have the benefit of a deliberate evaluation, even if the failure is delayed.

. The Staff proposes to leave the question of' seismic risk, including the fragility of equipment within the GESSAR II scope, to the construction permit stage. The Staff is confident that some, as yet unspecified, criteria for the seismic contribution to risk of severe accidents can be met at that stage without significant changes in the approved design. We do not share that confidence in the absence of a decision on a safety goal.

1 . The Staff consultants were ' provided only limited resources to review the internal flooding portion of the PRA. Because of the

limited effort and the unavailability of design details vital to an i evaluation of variout flooding scenarios, the consultants were not able to estimate adequately the flooding contribution to core melt, which the Applicant calculates to be small. Thus, while some j effort was made, the Staff's evaluation of the PRA was limited in j this respect.

Our concerns about the FDA process include chiefly two areas: (1) the l

amount of detail and completeness required for approval of a " final 8

Honorable Nunzio J. Palladino January 14, 1986 design" and (2) the nature and definition of the interfaces between the nuclear island and the balance of plant, especially those that must be expressed in terms of reliability to meet the intent of the Severe Accident Policy. These concerns are generic to the standard plant concept and have arisen in our deeper examination of GESSAR II in terms of the Severe Accident Policy. Their resolution necessarily will be evolutionary; but, in our opinion they have not been adequately resolved in the GESSAR II application and review. Some of our concerns are described more fully in the following:

. The Applicant has committed to incorporate an ultimate plant protection system (UPPS) in the GESSAR II design, which could reduce the incidence of core melt accidents. However, the detailed design of this system has not been provided; it is to be provided at the time of a specific plant application. As a result, the Staff has not been able to evaluate this proposed system, nor have we.

We are concerned that the scope of the FDA is not defined and  ;

documented with sufficient comprehensiveness and detail. We believe that this is necessary in order to make clear what changes in the design or in the plant can subsequently be required by the Staff without their being justified under the backfitting rule.

. The interface requirements are not sufficiently well specified in terms of minimum, quantitative performance requirements for systems and components of importance to an evaluation of core melt frequen-cy and risk. Hence, there is no real assurance that a plant built in accordance with the GESSAR II design will meet or better the Staff's estimates of accident frequency and consequences. Also, there is no interface requirement aimed at -limiting the number of challenges arising from the balance of plant to those assumed in the PRA.

Over and above the questions relating to the severe accident review and adequacy of the FSAR for an operating license stage document, the ACRS

! thinks that the following matters warrant consideration for the GESSAR

! II.

i l . We believe that the design of the scram discharge system has basic deficiencies in concept in the form of a preclosed dump volume.

Consideration should be given to means, which may be relatively simple, to avoid continuing problems with this design.

. We believe that there should be requirements for a study of the effects of seismically induced failures of nonseismically designed ,

components and structures on systems important to safety, for both '

i GESSAR II and the balance of plant. i

! l l

9

Honorable Nunzio J. Palladino January 14, 1986

. General Electric maintains that with their choice of materials and proper attention to water quality, GESSAR II should be essentially free of stress corrosion cracking. We do not believe that this can be assumed in view of the long prior history of surprises in regard to stress corrosion cracking. We recommend that any FDA should include provisions for monitoring and for replacement of deficient material.

Our findings and recommendations are as follows:

. We believe that the GESSAR II design includes features that have the potential to provide a significant improvement in safety over current BWR designs.

. We are unable to agree with the Staff, for reasons discussed previously, that the design satisfactorily or completely addresses all of the concerns described in the Comission's Severe Accident Policy Statement.

. We see no harm in the approval of the GESSAR II design, provided that this approval is for a limited time (say five years), and provided that this procedure not be viewed in any way as a prece-dent for the handling of future applications. In particular, the information provided to us in connection with GESSAR II would not be sufficient to support an application for a one-step license.

Additional comments by ACRS members Max W. Carbon and Charles J. Wylie and by ACRS Member David Okrent are presented below.

Sincerely,

. 4 David A. Ward Chairman Additional Comments by ACRS Members Max W. Carbon and Charles J. Wylie l It is our belief that the GESSAR II design represents an improvement in safety over BWR designs ap; met all NRC requirements.  ; roved

.Many itemsinremain the past andtothat open finalthe Applicant has resolution,  ;

but considerable additiona} review will be performed by both the Staff i

, and the ACRS for either ve- or two-step licensing. Therefore, we

support the Staff's plan t0 issue an FDA applicable to one-step licens-ing.  ;

. j

?

s

= d' 10

Honorable Nunzio J. Palladino January 14, 1986 l

Additional Comments by ACRS Member David Okrent I agree with the ACRS that the GESSAR II design (and NRC Staff review) does not satisfactorily or completely address all of the concerns described in the Commission's Severe Accident Policy Statement. I also agree with those specific concerns about the review and review process described in the ACRS report.

I do not concur with the Staff that the design and review are adequate for issuance of an FDA that has met the Severe Accident Policy State-ment, one which, according to the EDO reccmmended position, would be eligible for a five-year extension after a five-year initial award (and one for which the AIF proposes a ten-year approval period). I would have preferred rather that this be an interim report and that the entire matter, including the status of the GESSAR II review, be ciscussed by the ACRS with the Commissioners prior to further action on the GESSAR II FDA. In view of the multiple problems of inadequate design detail, incomplete Staff review, and potential conflicts with safety goal policy, among others, I do not think that GESSAR II should receive the

" qualified" FDA recommended by the ACRS at this time.

I would like to elaborate on some of the concerns raised in the ACRS letter and introduce others that are not mentioned in the ACRS letter, as follows:

1. The seismic design and seismic PRA are inadequately defined. In SSER No. 3, the Staff determined that the GESSAR II seismic risk study did not model well the risk likely to be contributed by seismic initiators for an actual GESSAR 11 plant at a typical site.

The Staff now reports that the point estimate seismic-induced core melt frequency might be as high as one-in-a-thousand per year for

" worst case" fragility values and unfavorable siting locations.

The Staff gives a point estimate of about 4-5 x 10-5 per year as the seismic contribution to core melt frequency, perhaps half of which is attributed to seismically induced relay chatter. The Staff's estimate of the seismic contribution to core melt frequency is not a mean value, and it is not practical to ascertain a mean from their reported results.

I am currently not able to ascribe a numerical value to the seismic contribution to risk. However, I believe that the Staff estimate of about 4-5 x 10-5 is too large to be accepted for the contribu-tion to core melt frequency from a single source or kind of acci-dent initiator. I believe this value is too large an overall core melt frequency to be accepted for a future plant or FDA. I recom-mend that an overall total large-scale core melt frequency with a mean value of 10-5 per year be taken as the objective for future plants, and that about one-fifth of this objective should be a somewhat flexible objective for any principal contributor, such as an earthquake. Limitations on the contribution from individual sources will help reduce the impact of large uncertainties.

11

Honorable Nunzio J. Palladino January 14, 1986 The Staff proposes to leave the question of seismic risk, including the fragility of components and equipment within the GESSAR II scope, to the construction permit stage. The Staff lists condi-tions to be met which could be interpreted as accepting a seismic core melt frequency such as the Staff estimates. The Staff further concludes that, if these conditions are not met, the utility applicant must demonstrate that this does not result in any significant increment in risk. But what is significant for a PRA?

Is it a factor of two? A factor of ten? The Staff provides no basis for judging what might be acceptable in this regard. The Staff also states that the site hazard curve must be bounded by the GESSAR II hazard curve, without explaining how uncertainties are to enter into such a bounding exercise.

Although GESSAR II is well into the design stage, I believe that the merits of probabilistic seismic design bases should receive consideration in trying to achieve a smaller contribution to overall risk.

2. I believe that the FDA should not be approved with such incomplete and sketchy information available for the proposed ultimate plant protection system (UPPS).

3. For future plants, I believe that a dedicated, safety-grade, independent system for removal of decay heat from the core and containment should be included, in consequence of the matters entering into the resolution of USI A-45, Shutdown Decay Heat Removal, unless a case can be made that all of its merits have been met in other ways. I favor hardening this system. This issue is discussed further in the next item.

4. In its review of GESSAR II, the Staff did not look beyond the current requirements for sabotage protection. In a letter to you dated July 17, 1985 concerning sabotage protection, the ACRS l recommended that the Commission reconsider its design basis threat definition for sabotage protection and decide if the present definition should be reconfirmed or modified. The Committee also recommended that the Commission consider whether the NRC Staff, in the course of reviews of new designs, should take account of design options, and possible combinations of measures, which might have j the effect of reducing or inhibiting sabotage or terrorist threats.

This matter should be dealt with before issuance of future FDAs, rather than as a possible backfit item. Specifically, I recommend that the following be factored into the design of GESSAR II (and ,

its balance of plant): l

. a protected, independent, safety-grade shutdown heat removal system 12 l

1 l

l Honorable Nunzio J. Palladino January 14, 1986

. protection of the control room and other vital areas or functions against a vehicle bomb at the edge of the guarded site periphery by proper location, building strength, or other measures

. geographical separation of redundant systems, including the ultimata heat removal system

. special monitoring and access control of especially sensitive protection systems 2 . roof design to limit helicopter landing access In sumary, I believe that neither the state of the design nor the 4

Staff's review process is adequate for issuance of a forward-looking FDA which has taken Severe Accident and Safety Goal Policy properly into account. This is particularly so in view of the Commission's own test in applying backfitting policy.

References:

1. General Electric Company Standard Safety Analysis Report, "GESSAR II, BWR/6 Nuclear Island Design," with Amendments 1 through 20

2. U. S. Nuclear Regulatory Comission, " Safety Evaluation Report

, Related to the Final Design Approval of the GESSAR II BWR/6 Nuclear Island Design" NUREG-0979, dated April 1983

3. Supplement 1 to the Safety Evaluation Report, dated July 1983

4. Supplement 2 to the Safety Evaluation Report, dated November 1984

5. Supplement 3 to the Safety Evaluation Report, dated January 1985

6. Supplement 4 to the Safety Evaluation Report, dated July 1985 i

e I

13 i

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,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 4.....

June 9, 1986 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555 J

SUBJECT:

ACRS RECOMMENDATIONS ON THE HOPE CREEK GENERATING STATION In its report to you of December 18, 1984 on the operating license appli-cation for the Hope Creek Generating Station, the Advisory Comittee on Reactor Safeguards made the following~ comments:

1) "Because of the nonoptimum orientation of the turbine relative to vital components in this plant, we recommend that a structured test program for evaluating overspeed protection of the turbine be prepared and submitted to the NRC Staff for review and approval before full power operation."

2) "Although the control room at the Hope Creek Generating Station has been reviewed with respect to human factors, we encourage the NRC Staff to give additional attention to its habitability requirements.

This should include evaluations of the potential loss of both trains of the emergency ventilation system and the heat load and rate of temperature rise in the room under a range of HVAC conditions."

Appendix H to Supplement No. 5 of the Hope Creek Generating Station Safety Evaluation Report includes the NRC Staff's response to both of these con-cerns.

With respect to the first, the Staff notes that the Applicant plans a loss-of-load test at 100 percent power which meets the requirements of Regulatory Guide 1.68, and observes that the requirements of GDC 4 are met. They also note, however, that the " test program proposed by the applicant does not actually test the trip setting of the mechanical and backup electrical overspeed trips, nor does it show that the trip setpoints (110% and 112% of rated set to speed limip the forturbine, mechanical and backup overspeed to 120% electrical of ratedoverspeed, respover-speed design (ectively) are speed)."

We do not consider this a basis for a conclusion that the turbine overspeed protection at Hope Creek is adequate, and suggest that the Staff be asked to provide justification based on safety considerations, prior to Commission

approval of full power operation.

15

i Honorable Nunzio J. Palladino June 9, 1986 i

The Committee's concerns about control room habitability in the event of loss of ventilation were addressed by an analysis. Although the Staff has con-cluded that a loss-of-ventilation and loss-of-cooling test would be necessary to completely address those concerns, the NRC Staff has concluded that nothing more is needed as part of the operating license review but that additional test requirements or limitations resulting from resolution of Generic Issue 83, " Control Room Habitability," would result in additional plant-specific action. We find this acceptable.

i Sincerely, p

,j .

i David A. Ward Chairman 1

1 16

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l p rero UNITED STATES 8 o,j NUCLEAR REGULATORY COMMISSION f

P a ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 July 15, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS RECOMMENDATIONS ON THE HOPE CREEK GENERATING STATION 1

In its report to Chainnan Palladino of December 18, 1984 on the operat-ing license application for the Hope Creek Generating Station, the Advisory Comittee on Reactor Safeguards made the following coment

"Because of the nonoptimum orientation of the turbine relative to vital components in this plant, we recomend that _a structured test program for evaluating overspeed protection of the turbine be prepared and submitted to the NRC Staff for review and approval before full power operation."

l During its 314th meeting, June 5-7, 1986, and after reviewing Appendix H

! to Supplement No. 5 of the Hope Creek Generating Station Safety Evalua-

tion Report, which included the NRC Staff's response to this concern, i the ACRS in its report to Chairman Palladino of June 9,1986, comented that the Staff's response does not constitute a basis for a conclusion that the turbine overspeed protection at Hope Creek is adequate. We also suggested that the Staff be asked to provide justification based on safety considerations, prior to Comission approval of full power

operation.

In view of new information provided by the Licensee, we conclude that the evaluation of turbine overspeed protection at the Hope Creek Gen-

erating Station is acceptable.

! Sincerely, I

David A. Ward Chairman i

17

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1. + UNITED STATES 8 'n NUCLEAR REGULATORY COMMISSION f

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. ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555 t ....*

April 16, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

PROPOSAL BY DUKE POWER COMPANY TO OPERATE THE MCGUIRE NUCLEAR STATION, UNITS 1 AND 2, WITHOUT UPPER HEAD INJECTION SYSTEMS During its 312th meeting, April 10-12, 1986, the Advisory Comittee on Reactor Safeguards reviewed the proposal by the Duke Power Company to 4 operate the McGuire Nuclear Station, Units 1 and 2, without the upper head injection system (UHIS) portions of the emergency core cooling systems. The ECCS Subcommittee met to review this proposal on February 21, 1985 and March 26, 1986. We heard presentations by the NRR Staff, which endorses the proposal, by Duke Power Company, and by Westinghouse Electric Corporation, the designer of the McGuire Nuclear Steam Supply System. The NRC Staff provided a preliminary Safety Aaluation Report on this topic. Their review bases were augmented dur1% discussion with us at this meeting.

The UHIS was added to the McGuire units and later added to most PWRs with ice condenser containments to improve core cooling during hypothe-l tical large break LOCAs. The UHIS includes a pressurized accumulator

, which provides a high volume flow of water directly into the upper head if the reactor coolant pressure decreases below approximately 1000 psig as would be expected in an LB-LOCA. This system supplements the ECCS accumulator, which provides high volume flow to the cold leg at lower pressure in most PWRs. Thus, following an LB-LOCA, the core will be refilled from top and bottom as opposed to only from the bottom as in most PWRs without the UHIS.

At the time of the McGuire operating license review, analyses of ECCS performance, using available " evaluation model" (EM) codes, indicated i

that the more rapid reflooding provided by UHIS was necessary to reduce i calculated peak clad temperatures (PCTs). The higher PCTs were the result of steam binding in parts of the reactor coolant system brought on by the lower containment back pressure typical of ice condenser containment performance during a hypothetical LB-LOCA. Without the UHIS, it would have been necessary to operate the McGuire units with re-strictions on core power or core power peaking. With the UHIS, permis-sible peaking factors were similar to those in other Westinghouse

plants.

Duke Power's experience with the UHIS at McGuire has not been entirely satisfactory. While there has been no indication that the system would 19 3

Mr. Victor Stello, Jr. not perform its function should an LB-LOCA occur, it has resulted in increased occupational radiation exposures and added to the complexity of plant operation and maintenance. There is strong incentive for its removal if it can be shown to be unnecessary.

Since the time of McGuire's operating licensing review, codes used for analysis of ECCS performance have-been im This includes both EM codes and those used to make (more nearly) proved.bestestimate(BE) analys Recent analyses, using the improved codes, by both the licensee and its contractors and the NRC Staff and its contractors, indicate that the performance of the McGuire ECCS, without the UHIS, is better than was

predicted by the codes of a decade ago. The licensee claims, and the

] NRC Staff concurs, that the calculated perfomance is enough improved that the McGuire units can safely be operated without the UHIS and with

~

normal core power limitations.

Using the improved EM codes has shown that licensing requirements can be met without the UHIS. There is also evidence, from the rather sketchy i BE analyses completed, that the UHIS does provide some benefit and that operation without UHIS will reduce real PCT margins. However, we are i persuaded that there are certain safety benefits in removing the UHIS, chiefly by reducing both the complexity of plant operation and its f

vulnerability to certain reactor upsets. We agree that the Duke Power Company proposal can be accepted. However, we suggest that they con-sider maintaining the removal and disconnection of the UHIS on a re-versible basis until further calculations show that the core operating conservatisms of McGuire without UHIS are similar to non-UHIS units.

We believe our review of this matter points up the need to expedite revision of both 10 CFR 50.46 and Appendix K. The reliance on overly conservative evaluation models as mandated by the above regulations adversely impacts nuclear power plant safety. ,

Mr. C. J. Wylie did not participate in the Comittee's review of this matter.

Sincerely,

. N .

David A. Ward Chairman

Reference:

Memorandum dated March 18, 1986 from C. H. Berlinger, NRR, to P.

Boehnert, ACRS Staff,

Subject:

Draft SER for the Proposed Operation of McGuire Units 1 and 2 Without Upper Head Injection, with enclosures 20

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l 8 o UNITED STATES '

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8" o NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$  ;, W ASHINGTON, D. C. 20555 g ,g

! August 12, 1986 i

i MEMORANDUM FOR: Victor Stello, Jr. i Executive Director for Operations FROM: R. F. Fra cutiv irector, ACRS

SUBJECT:

PROF 0 SED POWER INCREASE FOR NORTH ANNA POWER STATION, UNITS 1 AND 2 Virginia Electric and Power Company (Licensee) filed an application for an amendment dated May 2, 1985 to operate the North Anna Power Station, Units 1 and 2 at an increased power level from 2775 MWt to 2893 MWt.

This proposed change, which represents approximately a 4.2% power incrcase over the presently licensed core power, was brought to the attention of the Advisory Committee on Reactor Safeguards on August 9, 1985, but no action was taken by the Committee.

1 The North Anna Updated Final Safety Analysis Report (UFSAR) notes that the environmental analyses and accident evaluations were performed at a stretch core power rating of 2900 MWt and an NSSS power rating of 2910 MWt. I understand that the NRC Staff anticipates approval of this power increase by mid-August 1986.

During its 316th meeting, August 7-9, 1986, the ACRS concluded that this power level increase does not warrant review by the Conunittee, cc:

H. Denton, NRR T. Novak, NRR L. S. Rubenstein, NRR L. B. Engle, NRR R. Hernan, NRR I

21

1. UNITED STATES

! o,f, NUCLEAR REGULATORY COMMISSION

$ f ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wasumarow.o.c.aoses November 12, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON NUREG-1206, " ANALYSIS OF FRENCH (PALUEL)

' PRESSURIZED WATER REACTOR DESIGN DIFFERENCES COMPARED TO CURRENT U.S. PWR DESIGNS" DATED JUNE 1986 During its 319th meeting, November 6-8, 1986, the Advisory Committee on Reactor Safeguards was briefed by members of the NRC Staff concerning the Staff,'swater pressurized review of the reactor safety)

(Paluel and differences a typical U.S. between four-loopaPWR typical-- aFrench SNUPPS plant. A subcommittee meeting was previously held to discuss this subject on September 25, 1986. NUREG-1206, which has been pub-lished by the Staff, is an excellent report on this subject.

The NRC Staff had previously perfomed a review of the safety differ-ences between the British Sizewell B design and the SNUPPS plants.

These efforts were in part pursuant to an ACRS recommendation that the Staff perfom such studies of the differences in safety design between U.S. plants and those in other countries.

We believe that these efforts by the Staff have been worthwhile and should be continued. The Staff representatives suggested that the KONVOI series of the Federal Republic of Germany (FRG) might be a good next choice to study. We agree that KONVOI is a good candidate, and suggest that the comparison might involve a more recent U.S. design such as the South Texas project. We would also like to see a boiling water reactor (FRG, Swedish, or a Japanese design) compared with U.S. BWRs.

In any event, the ACRS recomends continuation of this program.

Sincerely, David A. Ward Chaiman 23

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4 smtuq#o UNITED STATES 8 g NUCLEAR REGULATORY COMMISSION p ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ WASHINGTON, D. C. 20665 March 17, 1986 i

4

Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D. C. 20555 j

SUBJECT:

ACRS REPORT ON THE PERRY NUCLEAR POWER PLANT, JNIT 1 During its 311th meeting, March 13-15, 1986, the Advisory Comittee on Reactor Safeguards discussed the Ohio earthquake which occurred on i January 31, 1986 and reviewed its implications with respect to the Perry Nuclear Power Plant, Unit 1. A meeting of the Subcommittee on Extreme

?

External Phenomena was held on March 12, 1986 to consider this matter.

l The Reactor Operations Subcommittee discussed this matter during a i meeting on February 12, 1986 and the full Comittee was briefed during i the 310th ACRS meeting, February 13-15, 1986. During our review, we had the benefit of discussions with representatives of Tt.e Cleveland Elec-tric Illuminating Company (Applicant), the NRC Staff, and the U. S.

Geological Survey (USGS) Staff. We also had the benefit of the docu-We com-ments referenced and comments from several ACRS consultants.

' mented on the application for a license to operate this plant on July i 13, 1982.

This earthquake occurred near Leroy, Ohio, and was characterized by relatively low energy, low velocities, small displacements, a short duration, and a response spectrum rich in high frequencies. Except at the relatively less significant high frequencies, the excitation of the

.j plant structures and equipment was much less than that considered in the i

seisr.ic design basis.

No significant damage was observed at the Perry plant in the inspections ,

j which were performed by the Applicant and the NRC Staff. The Cleveland Electric Illuminating Company, by usIng analysis and comparisons with i

prior qualification testing, has found that all the structures and equipment analyzed thus far have substantial margins of safety relative l

to the loads and stresses induced by the earthquake. The NRC Staff and j our consultants concur in that conclusion.

2 The NRC Staff has several confirmatory actions that it will require of the Applicant prior to operation above 5 percent power. These con-

! firmatory actions include the analyses of a large sample of equipment.

We support the NRC Staff's proposed confirmatory actions.

There currently exists some possibility that the January 31, 1986

' earthquake is related to deep well injection activities that took place between the Perry phnt site ar.d tht. town of Leroy or to solution mining 1

i 25

Honorable Nunzio J. Palladino March 17, 1986 that took place in this area. The NRC Staff has engaged the services of USGS to evaluate these hypotheses to see if there really may be a causal connection, and, if so, whether there is any likelihood of substantially larger earthquakes in the future. The NRC Staff will keep the ACRS informed as to the progress of the USGS work.

One of the ACRS consultants suggested that monitoring with sensitive seismological instruments over the next few years would be helpful in assessing the possible causal connection between the deep well injection and the January 31, 1986 earthquake. The USGS representatives attending our discussions agreed that such seismic monitoring would be valuable.

Therefore, unless the USGS and the NRC are able to decide that there is no causal connection or that earthquakes of a magnitude sufficient to be of concern can be ruled out from this cause, we reconsnend that The Cleveland Electric Illuminating Company assure that appropriately sensitive monitoring be continued over the next few years.

We agree with the NRC Staff that the January 31, 1986 earthquake is unlikely to lead to any requirements that would significantly change the design of the Perry plant's structure or its equipment. Based on the infomation developed in these meetings and considering the above consnents, we find no reason to alter the conclusions stated in the Consni ttee 's report dated July 13, 1982 regarding operation of this nuclear plant.

Sincerely, e h David A. Ward Chairman  !

References:

1. The Cleveland Electric Illuminating Company, " Seismic Event Evaluation Report, Perry Nuclear Power Plant," dated February 1986

2. Letter dated March 5, 1986 from Robert M. Bernero. Director, Division of BWR Licensing, NRC, to David Ward, Chairman, ACRS,

Subject:

Perry Seismic Safety Evaluation, with attached Safety Evaluation Report dated March 1986 i

)

26 l

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8 n NUCLEAR REGULATORY COMMISSION

,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20$55 Q

June 10, 1986 l Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS REPORT ON SOUTH TEXAS PROJECT, UNITS 1 AND 2 During its 314th meeting, June 5-7, 1986, the Advisory Committee on Reactor Safeguards reviewed the application of Houston Lighting and Power Company (HL&P), the Applicant, acting on behalf of itself and as agent for the City Public Service Board of San Antonio, Central Power j and Light Company, and City of Austin for a license to operate the South Texas Project, Units 1 and 2. The ACRS contented on the construction permit application for the South Texas Project, Units 1 and 2 in a report dated September 19, 1975. The ACRS Subcommittee on the South

]

Texas Project toured the facility on May 29, 1986 and met in Bay City.

Texas on May 29 and 30,1986 to discuss the application. During our review, we had the benefit of discussions with ' representatives and consultants of the Applicant Westinghouse Electric Corporation. Bechtel Energy Corporation, and the NRC Staff. We also had the benefit of the documents referenced.

The site is located in south-central Matagorda County west of the Colorado River, 8 miles north-northwest of the town of Matagorda and about 89 miles southwest of Houston. The plant is located about 12 miles south-southwest of Bay City. Westinghouse Electric Corporation is the nuclear steam supply system and turbine-generator supplier for South Texas Project, Units 1 and 2. This Project makes use of identical four-loop Westinghouse pressurized water reactors and turbine generators.

Unit 2 is similar to Unit I and is 600 feet away. This is the only U.S.

plant using the RESAR-41 design. Although this design differs in some respects from other Westinghouse four-loop units in this country, it is quite similar to the Paluel plant in France, which is now in operation.

Unit 1 is approximately 90 percent complete, and it is scheduled to load fuel in June 1987. Unit 2 is expected to follow about eighteen months later. The Applicant appears to have assembled a capable and experi-encedytaff.

During our meeting, the NRC Staff identified a number of issues that must be resolved prior to the granting of an operating license. The residual heat removal pump is located inside containment. While this 4

offers some advantages, it will be necessary that the pump be qualified for operation in an accident environment before this system can be judged acceptable. We wish to be kept informed.

27 1

1 Honorable Nunzio J. Palladino June 10, 1986 s

i 4

We heard a report from a representative of the NRC's Region IV Office J that construction quality and quality assurance effectiveness at the

. South Texas Project were satisfactory and that the attention being given 4

by management to all aspects of the plant's readiness was comendable.

However, the results of a recent Construction Appraisal Team inspection j which are presently being considered may introduce items requiring i attention.

l In its report of September 19, 1975 on the construction permit appli-i cation, the ACRS asked to be kept infonned on the resolution of several j items, including the location of the storage tanks for the diesel fuel.

The diesel fuel storage tanks are located in separate rooms above the diesel generators. . With this arrangement, a major concern is that a i break in the piping between tne storage tanks and the diesel generators .

will result in an uncontrolled discharge of fuel oil which may cause a l fire. The ACRS recommends that the Applicant perform tests and take i t

appropriate corrective measures to prevent failures in fuel oil piping and tubing by induced vibration resulting from extended operation of the

diesel generators.

4 We believe that, subject to the resolution of open items identified by

, the NRC Staff and the items noted above, there is reasonable assurance

, that the South Texas Project, Units 1 and 2 can be operated at power levels up to 3800 Mwt without undue risk to the health and safety of the l public.

Sincerely, J

. i i

David A. Ward

{ Chairman i

References:

1. Final Safety Analysis Report for South Texas Project, Units 1 and a 2, Volumes 1-16, including Amendments 1-53 t

2. U.S. Nuclear Regulatory Connission, " Safety Evaluation Report

Related to the Operation of South Texas Project, Units 1 and 2,"

USNRC Report NUREG-0781 dated April 1986 i

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28

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• o UNITED STATES 8" n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS E

$ , W ASHINGTON, D. C. 20655 e g v,***** # February 18, 1986 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS REPORT ON PROTECTION AGAINST CRITICALITY DURING TMI-2 DEFUELING During its 310th meeting, February 13-15, 1986, the Advisory Comittee on Reactor Safeguards met with representatives of GPU Nuclear and the NRC Staff to review studies done by GPU Nuclear and procedures proposed to protect against criticality during the defueling of the TMI-2 core.

The Core Performance Subcommittee had previously reviewed this matter during their meeting on January 29, 1986.

We conclude that appropriate studies have been performed and that the procedures being followed will provide the necessary protection against criticality during defueling.

Sincerely, David A. Ward Chairman

References:

1. Transcript of meeting between U. S. Nuclear Regulatory Comission and Advisory Panel on Decontamination of TMI-2 on November 19, 1985

2. Letter from F. R. Standerfer, GPU Nuclear, to B. Snyder, NRC, dated November 8, 1984

Subject:

Reactor Coolant System Criticality Report

3. Memorandum from C. Berlinger, NRC, to R. Weller, NRC, dated February 13, 1985,

Subject:

Review of TMI-2 Reactor Coolant System Criticality Report

4. Letter from F. R. Standerfer, GPU Nuclear, to B. Snyder, NRC, dated September 27, 1985,

Subject:

Revision 2 to Boron Hazards Analysis

5. Memorandum from L. Thonus, NRC, to W. Travers, NRC, dated October 25, 1985,

Subject:

Boron Dilution Hazards Analysis, Revision 2

6. Memorandum from C. Berlinger, NRC, to R. Weller, NRC, dated Decem-ber 2, 1985,

Subject:

Review of TMI-2 Foreign Materials Criti-cality Report l

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! Part 2: ACRS Reports on Generic Subjects

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8 n NUCLEAR REGULATORY COMMISSION

,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 8 WASHINGTON, O. C. 20665 o

• • April 16, 1986 1

Honorable Nunzio J. Palladino

. Chairman j U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS REGARDING NRC REVIEW OF ADVANCED REACTOR j DESIGNS

, During its 311th and 312th meetings, March 13-15 and April 10-12, 1986, the Advisory Committee on Reactor Safeguards heard presentations by the NRR Staff, DOE personnel, and DOE industrial subcontractors on i one advanced gas-cooled reactor (GCR) design and two advanced liquid-i metal reactor (LMR) designs. These designs are in their early stages,

and a unique feature of the design efforts is that NRR personnel have l

provided safety input very early in the conceptual design stage. This

approach, which is in accord with the NRC Advanced Reactor Policy Statement, contrasts with that followed in the design of most of the current generation light water reactors (LWRs) wherein a finalized design was presented to NRC for review and approval (or disapproval).

The ACRS believes that significant safety benefits can result from an j early interaction between the NRC and the designers and that NRC can have a fundamental influence on the safety aspects of a design if its

input is provided at an early stage when design changes can be made both easily and without substantive cost. This contrasts with the situation wherein a finished design is presented to NRC and the latter has considerable difficulty influencing the safety design of the reactor other than through " patches" or " add ons " as some have described the process. The ACRS has recomended the early-interaction approach in the past, and we continue to support it strongly.

! These design efforts are directed toward achieving high levels of safety as well as toward achieving low costs and improved operating <

features. They are thus aimed toward implementing the policy of the i

Congress as expressed in the Atomic Energy Act. Many innovative features are evolving. For example: l r

1. LMR designs are being developed which the designers believe would ,

j tolerate, without core melt or significant radiation release, '

! very severe accidents such as loss of flow without scram, power excursion without scram (both commonly referred to as ATWS for LWRs), and loss of heat sink without scram. These designs are i being influenced by tests run during the past months on EBR-II in Idaho, which have proved that some LMRs can indeed tolerate such severe accidents without public health effects.

31

Honorable Nunzio J. Palladino April 16, 1986

2. The designers believe that the need for. emergency evacuation planning for the surrounding population can be totally or almost totally eliminated.

3. The reactors which are evolving are small, modular units that would be built in a central factory and shipped by truck, rail, or barge to a site. With factory fabrication, it should be possible to eliminate most of the QA/QC problems which have harassed the current LWRs. With small units, the capital costs per unit should be low, a feature attractive to prospective purchasers.

4. Designs may evolve for which no operator actions would be re-quired in the case of some severe accidents, fires, or types of sabotage for at least several hours.

These and many more innovative features are evolving. However, in order to optimize a design, it may not be necessary to incorporate safety features which would be required in a current LWR. The design-ers believe that they cannot be innovative in selected areas only; they believe they must be innovative across the board if they are to succeed.

We have been told by NRR Staff that their budget is being reduced drastically and that it may be necessary to terminate the early interactions with DOE. We are also told by DOE that it will be a great loss if this interaction ceases, that DOE and its subcontractors will be unable to proceed effectively without NRC safety input and regulatory guidance. Further, DOE will probably need to share costs with industry, and the latter may be more inclined to provide fi-nancial support if DOE can make some sort of statement that NRC considers the designs to be licensable.

We believe that it would be very shortsighted for NRC to terminate this effort for budgetary reasons. We realize that the agency has severe financial problems, but the total amount of resources involved here is very small, and we strongly urge a continuation of this modest effort. If DOE proceeds without NRC input, the NRC may have missed a golden opportunity to influence reactor safety. If DOE stops, the NRC may bear part of the responsibility for failure of the Congressional policy.

Although the comments above have been based on GCR and LMR activities which have been before us recently, the underlying considerations 32

Honorable Nunzio J. Palladino April 16, 1986 l

pertain fully as much and perhaps even more to advanced LWRs now being developed and designed by various U.S. organizations.

Sincerely, i s i Y.

David A. Ward Chairman l

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asaw o UNITED STATES g, g 3 o NUCLEAR REGULATORY COMMISSION f, E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS Q, g WASHINGTON, D. C. 20555 4*****$ March 18, 1986 i

l Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON THE IMPLEMENTATION PLAN FOR THE SEVERE ACCI-DENT POLICY STATEMENT AND REGULATORY USE OF NEW SOURCE-TERM INFORMATION During its 311th meeting, March 13-15, 1986, the ACRS discussed with the NRC Staff a draft of the Implementation Plan which is being prepared in accordance with the Commission's Severe Accident Policy Statement. This proposal was also discussed with the Class 9 Accidents Subcommittee on February 24, 1986.

The NRC Staff proposes to separate implementation into three somewhat independent areas: (1) systematic evaluation, on an individual basis, of plants now in operation or under construction; (2) analysis of proposed new plants using a combination of deterministic and probabilis-tic criteria; and (3) changes in regulations based on new source tenn information. A major part of the discussion was based on the first area which is more developed than the other two.

Systematic Evaluation of Individual Plants A procedure for systematic evaluation of operating plants is being developed in cooperation with those responsible for the IDCOR program.

It is to be based on insights gained from risk analysis but will be '

composed of a set of deterministic guidelines and criteria to be used by the staff of each operating plant. The general approach has been agreed upon by the NRC Staff and members of the IDCOR organization. However, a significant number of issues on which resolution is yet to be reached remain. It is proposed that the systematic evaluation treat so-called internal accident initiators, as well as internal flooding and internal fires. However, the approach to be used for dealing with seismic events and some other external accident initiators is yet to be developed.

We believe the proposed approach for individual plant analysis is reasonable. We expect to follow the progress of issue resolution and the results of a trial application of the method to several plants.

Although we expect that the problem of dealing with seismic initiators will be difficult, we believe it is important that the issue be 35

i i

Mr. Victor Stello, Jr. March 18, 1986 resolved. We urge that the difficulty not lead to postponement of the effort.

We note that one of the principal purposes of the individual analysis is to identify outliers that may be characteristic of individual plants.

We recommend that continuing attention be given to this purpose. We are not convinced that it will be achieved automatically by the proposed '

procedure. If it is to be achieved, those carrying out the analysis will, at the very least, need to be enthusiastic about the value of the I process and its results. The NRC Staff should make an effort to form-ulate a procedure that is perceived by those who will use it as having the possibility of enhancing plant safety with reasonable effort. It is

especially important that diligence in carrying out the search for vulnerabilities not lead to undue penalties.

We also recommend that the program for correcting any plant deficiencies discovered in this analysis or in the . course of resolving Unresolved Safety Issues be integrated in such a way that the proposed changes in plant equipment, in procedures, or in staffing are dealt with on a plantwide basis in order that they not conflict one with the other.

Analysis of Proposed New Plants Insights gained from NUREG-1150, " Nuclear Power Plant Risks and Regula-

tory Applications," and from other sources of information, will be used

! to develop some combination of deterministic requirements and probabi-i listically based criteria for judging plant response to severe' acci-l dents. The NRC Staff is developing requirements for the acceptable

! content of PRAs required for the licensing of new plants and criteria l for the interpretation of and the regulatory application of the PRA results. We have previously suggested that development of some com-

, bination of deterministic and probabilistic criteria might be desirable.

i We reserve further comment until the detailed approach-has been devel-

, oped.

' We recognize the current program to explore development of containment criteria. We note in NRC Staff comments on this program that a decision is to be made as to whether containment criteria are needed. Unless a  !

i decision is made to eliminate containment, there is no question of whether criteria should exist but only of what they are to be. The 3

present criteria were developed from a consideration of Design Basis Accidents. Because of the importance of containment performance in severe accidents, we are convinced that containment criteria should consider severe accidents as well.

Source Tenn Related Changes in Regulations i

! In our meeting, several regulatory areas were described which are to be l examined in the light of new source tenn information. We have repeat- j edly recomended that such an examination be made in order to guide the l

1 - , - - - . -

i Mr. Victor Stello, Jr. March 18, 1986 i

I severe accident research program. We have no quarrel with the areas chosen. We will comment further after the analyses have been completed.

As a general coment, we observe that the first two areas of implementa-tion activity seem to be schedule-driven to an extent that makes it possible that the decisions reached will suffer from incomplete study and developmert. We urge that adequate time be given to the treatment of these important issues.

Sincerely, t

s \

David A. Ward Chairman References

1. U. S. Nuclear Regulatory Commission, "NRC Policy on Future Reactor Designs - Decisions on Severe Accident Issues in Nuclear Power Plant Regulation," USNRC Report NUREG-1070, dated July 1985

2. SECY-86-76 (Draf t), " Implementation Plan for the Severe Accident Policy Statement and the Regulatory Use of New Source-Term Informa-tion," memo for the Commissioners from V. Stello, Acting Executive Director of Operations, dated February 28, 1986 l

W 37

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1. p urg'og UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

{, ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ ,g WASHINGTON, D. C. 20555

• • • June 10, 1986 l

Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON NUREG-0956, " REASSESSMENT OF THE TECHNICAL BASES FOR ESTIMATING SOURCE TERMS" -- REVIEW COPY During its 314th meeting, June 5-7, 1986, the Advisory Comittee on Reactor Safeguards discussed NUREG-0956 with representatives of the NRC Staff. This report had previously been reviewed by a Subcommittee in a meeting on June 3, 1986. A draft copy of this report issued for public coment had been reviewed by the Committee during its 306th meeting, October 10-12, 1985, and an ACRS report was issued on December 12, 1985.

We also had the benefit of the documents referenced.

In our letter of December 12, 1985, we made a number of coments on the draft report. Our review of the revised version, on which we report in this letter, indicates that a number of changes have been made. We consider this version to be superior to the one we reviewed earlier. We observe, however, that:

(1) Although a variety of uncertainties are associated with both the physical phenomena and the calculational tools described in NUREG-0956, the description of the uncertainties given in the report is only qualitative. In order to use the material in this report, and to draw conclusions from the results, information on uncertainties

must be available. We are told that proposed NUREG-1150, " Nuclear Power Plant Risk and Regulatory Applications," will contain what-ever quantitative description of uncertainties can be developed with existing information.

Whether this information will be presented in a way which will make it possible to identify the uncertainties attributable to the Source Tenn Code Package (STCP) is not clear. If the information in NUREG-0956 is to be generally useful, this identification is essential. We recommend that specific attention be given to this identification,*either in NUREG-0956 or in proposed NUREG-1150.

(2) In view of the variety of earlier coments indicating that source term research would provide significant new infonnation, surpris-ingly little is said in the report about the implications of the 39

Honorable Nunzio J. Palladino June 10, 1986 new information that has been developed. The report claims signif-icant improvements in calculational methods compared to earlier work in WASH-1400, " Reactor Safety Study -

An Assessment of Accident Risks in U.S. Commercial Nuclear Power Plants." But about all that is said concerning the results of the new methods is that one cannot generalize from them. Section 4.13 of NUREG-0956 does describe a comparison of iodine species. More of this kind of information would make the report more useful. We note, however, that a recent NRC Report, NUREG-1171, " Draft Environmental State-ment Related to the Operation of South Texas Project, Units 1 and 2," does directly compare both early and delayed fatalities as calculated by the "old" and the "new" source terms.

(3) So far as we can determine, the use of the STCP requires that many decisions that have a significant influence on the calculated core melt progression must be made by the user. User decisions thus have a significant influence on sequence consequences. Experienced users are aware of this, but the report should point this out for the benefit of those less familiar with the package.

(4) It would be helpful to identify the methodology of source term calculations independently of the particular codes that make up the current STCP. For example, an appendix describing the hand calcu-lations that form the basis for the infonnation in Table 3.4 of NUREG-0956 would be useful.

(5) We reiterate the importance of complete and accurate documentation.

This is especially critical in this report because of the many documents on which the report depends. The version of the report which we reviewed had a considerable number of errors and omis-sions. We assume these will be corrected.

(6) There are still a number of obvious deficiencies in the physical modeling of the codes. For example, in-vessel circulation and ex-vessel time-dependent release of molten core material are not  !

treated. j (7) An early goal of the source term research program, and the subse-quent recalculation of risks for several representative power i plants, was the formulation of a generic source term that would j permit an estimate of the risk produced by most of the reactors now operating. This report hints that the development of a generic soprce term may pot be feasible. If a less ambiguous statement can be made, it could be useful to those responsible for future re-  !

search and regulatory changes. j l

(8) An impor tant consideration in risk-impact studies is the biological  ;

significane' of the nuclides that make up the source term. A table giving this information would be helpful to those attempting to understand the risk significance of the source tenn information reported.

40

Honorable Nunzio J. Palladino June 10, 1986 l

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(9) The report refers in several places to hydrogen burns. The report does not make clear when a burn is assumed to happen. In many situations there is probably enough steam to inert the atmosphere.

In addition, the core-concrete intert tion produces more H,0 and C07 than H2 Plus C0. While this would not create an inert ' atmos-phere when injected into a large volore of clean air, it would become inert after one or more burns of 8% (H, + C0), with the resultant H 2O and C02 . Whatever assumptions are used should be made clear (10) We find it remarkable that no serious effort has yet been made to model the TMI-2 accident.

Sincerely, d

.I David A. Ward Chairman

References:

1. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, " Reassessment of the Technical Bases for Estimating i Source Tenns," Review Copy, USNRC Report NUREG-0956, dated May 23, 1986

2. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Regulation, " Draft Environmental Statement Related to the Operation of South Texas Project, Units 1 and 2," USNRC Report NUREG-1171, dated March 1986

3. U.S. Nuclear Regulatory Commission, " Reactor Safety Study - An 1.

Assessment of Accident Risks in U.S. Comercial Nuclear Power Plants," USNRC Report WASH-1400 (NUREG-75/104), dated October 1975 1

l 41

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~g UNITED STATES E o NUCLEAR REGULATORY COMMISSION

$ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

.I WASHINGTON, D. C. 20555 1  %,...../ December 17, 1986 i

i l

The Honorable Lando W. Zech, Jr.

I Chairman i U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Chaiman Zech:

l

SUBJECT:

ACRS COMMENTS ON PROPOSED BWR MARK I CONTAINMENT REQUIREMENTS j FOR SEVERE ACCIDENTS j During our 320th meeting, December 11-13, 1986, the members of the i Advisory Committee on Reactor Safeguards discussed proposed new require-I ments for Boiling Water Reactor (BWR) Mark I containments with regard to i their capability to withstand severe accidents. During our review, we j had the benefit of discussions with representatives of the NRC Staff and BWR Owners' Group (BWROG), as well as the benefit of the documents referenced. Discussions were also held with the Staff during our 315th meeting, July 10-12, 1986, and information was developed during the

, Containment Requirements Subcommittee meetings on September 23 and 1

December 9,1986.

! The Staff's proposed requirements for BWR Mark I containments were presented in a draft generic letter with an attachment evaluating containment performance during severe accidents. This evaluation is admittedly open to question ir regard to conditional containment failure i probabilities, and it lacks detailed technical justification for the se-lection of certain procedures or parameters, e.g., a reduction of 90% in i the drywell spray flow rates. We recommend that the Staff strengthen, to the extent feasible, the detailed technical analysis to support the proposed requirements.

l We also believe that additional information is needed in several areas,

! including the following:

e an estimate of the contribution to core melt frequency and to i containment failure from significant external events, including ,

I seismic events greater than the Safe Shutdown Earthquake

!' e a tabulation of information concerning significant differences among the family of Mark I containments sufficient to ascertain I that the proposed generic requirements would not necessitate l special exceptions and/or additions r

43

The Honorable Lando W. Zech, Jr. December 17, 1986 e an examination of possible adverse effects of pool bypass as a result of transient thermal stresses and their possible effect on j drywell connections to downcomers I

e an analysis of the proposed Emergency Procedure Guidelines (for l example, circumstances under which emergency sprays may be initi- )

ated or reinitiated) which considers the effects of venting and spraying operations on the continuing availability of the vapor suppression processes and other engineered safety features which may continue to be needed or highly desirable l Until sufficient information has been developed on matters such as these, we are unwilling to agree with the proposed position in the draft generic letter: "Given the implementation of the generic improvements i of Mark I containments, there is no need for an Individual Plant Eval-uation (IPE) for containment performance."

Nevertheless, the Staff should issue the proposed generic letter for public comment. The BWROG containment studies and the Staff's evalua-tion of the Emergency Procedure Guidelines are expected to be issued during the proposed public comment period. The results of these studies should be considered along with public comments as part of the final resolution of this issue. We would like to consider this matter at a future meeting when the actions noted above are completed.

Sincerely, d

David A. Ward Chairman

References:

i

1. Memorandum from R. Bernero for R. Fraley,

Subject:

Proposed Generic letter on Improvements for BWR Mark I Containments, dated '

a December 3, 1986.

2. BWR Owners' Group, " Emergency Precedure Guidelines," OEI Document 8390-4, Draft Revision 4AF, dated August 14, 1986.

j 44 4

l l  ! ossa ~

o UNITED STATES l  ! o NUCLEAR REGULATORY COMMISSION I

{  ; ,E g

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

$9 * * * * * ,o' July 16,1986 Mr. Victor Stello, Jr.

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

SUBJECT:

EMERGENCY PLANNING REQUIREMENTS FOR NRC LICENSEES The ACRS has noted that the Consnission is considering requiring fuel cycle and other radioactive material licensees either to develop plans for coping with accidental radionuclide releases or to justify the reasons that the development of such plans is not needed.

To minimize the impact of this decision, we suggest that the NRC con-sider recommending that such licensees utilize in this process the screening model recently published by the National Council on Radiation Protection and Measurements. Such an approach could minimize the work involved and would enable licensees to apply a method that has undergone careful review and has been shown to provide a conservative estimate of the risks involved due to atmospheric radionuclide releases from a wide range of nuclear operations. Details of the screening model have been published as follows:

" Screening Techniques for Detennining Compliance With Environ-mental Standards," Commentary No. 3, National Council on Radiation Protection and Measurements, Bethesda, Maryland (March 31,1986).

' Sincerely, P

i I David A. Ward Chairman 45

UNITED STATES

1. % NUCLEAR REGULATORY COMMISSION 8 i

{ I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsHmoTow o.c.zosos g

November 10, 1986 j

Honorable Lando W. Zech, Jr.

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

SUBJECT:

PROTECTIVE ACTION GUIDES FOR NUCLEAR EMERGENCY PLANNING Included among the factors used by the NRC Staff in its evaluation of the required sizes of the Emergency Planning Zones around nuclear power plants are the Protective Action Guides (PAGs) recomended by the U.S. Environmental i ProtectionAgency(EPA).

During the time period since these guides were prepared, the concept of the-effective dose equivalent has been developed by the International Comission on Radiological Protection, and substantial changes have. been made in the thyroid dose considered to be equivalent to a given dose to.the whole body.

For these reasons, we believe that an updating of these guides would be useful. We understand that EPA is working on such an updating as part of its preparation of a broader set of PAGs. We suggest that the NRC encourage the EPA to proceed with the development of the new Guides in a timely manner and that the results be made available promptly for use in emergency planning evaluations.

Sincerely,

.1 David A. Ward Chairman 47

1. UNITED STATES d 'o

~g l 8" ,E o NUCLEAR REGULATORY COMMISSION l f o, g ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555 l

0,,,4 July 16, 1986 Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS VIEWS ON FIRE PROTECTION RESEARCH AND FIRE-RELATED SYSTEMS INTERACTIONS During its 315th meeting, July 10-12, 1986, the Advisory Connittee on Reactor Safeguards considered various activities associated with nuclear power plant fire protection provisions, including the Appendix R modi-fication process and fire-related portions of the safety research program. These matters were also discussed by the ACRS Subcommittee on Auxiliary Systems at a meeting held on June 26, 1986. We identified specific items of concern in the safety research area. We would like to share these since they may be of interest to you, and as a basis for reconsideration of resource allocation for fire-related research.

It is our understanding that all fire-related safety research programs will be terminated at various stages of completion by the end of FY 1986. In FY 1987, the Office of Nuclear Regulatory Research intends to perform a scoping study to reexamine the program needs, to determine the likely level of fire risk, and to establish future priorities. Our first concern is the loss of program momentum and information that will result from premature termination of ongoing activities while awaiting resul ts of the proposed scoping study. Any termination of needed research is a loss, and wi.ll be costly to reconstitute later.

The NRC Staff indicates that, based on probabilistic Risk Analyses (PRAs) that have been performed, there is a significant contribution to core-melt frequency from fire in a number of plants. Although the risk associated with fire is subject to large uncertainties, it is clearly nontrivial and the ability of fire to exacerbate severe accident situa-tions has already been demonstrated. Therefore, one needs to understand better how fires can start and propagate, and the consequential effects of fire and fire-fighting measures (or their inadvertent actuation) on safety-related equipment throughout the plant. Such information needs to be developed if the risk is to be properly analyzed and minimized, and the value-impact of potential backfits appropriately calculated.

The uncertainty must be narrowed if the results are to be useful for decision making.

In this regard, fires in the control room area are of special concern because of the confluence of essential plant control systems and the 49

4 The Honorable Lando W. Zech, Jr. July 16, 1986 likely sensitivity of control room components to the adverse environment i resulting from fire and fire suppression. Although considerable re- l search has been done in this area, important work remains to be complet-ed and documented concerning fire source charactr.rization, full-scale room environment tests, component failure threshe'id tests, and valida-tion of fire environment computer codes against full-scale room tests.

A majority of this work for FY 1986-1987 will remain unfinished or undone. The fire PRAs performed to date have not had the benefit of sufficient test data and information such as would be developed by this work. This significantly increases the uncertainty in the risk esti-mates. In addition, much of this work would add to a technical base that would be applicable to the analysis of fire situations, risk aversion, and value-impact considerations throughout the plant.

Our second concern relates to the fire-induced control system inter-actions issue which we thought was going to be within the scope of the resolution of USI A-17. " Systems Interactions in Nuclear Power Plants."

J We are now aware that the proposed resolution of USI A-17 does not include fire-induced interactions or those created by fire mitigation efforts or the effects of a seismic event on fire mitigation features.

We are concerned that this aspect of the issue has been overlooked. A research effort may be required to explore adequately the issue and its ramifications, including its potential contribution to connon cause and

. severe accidents. The results of some of the previous and proposed

! (but, now terminated) fire-related research may be needed in support of this effort.

We urge reconsideration of the budget and manpower allocations that have

led to the termination of the fire-related portions of the safety research program, and would like to be kept informed of how the fire-i induced system interaction issue will be handled. We are anxious to see a continuity of effort on fire-related research while the Staff deter-mines its future priorities.

! Finally, we believe that if the NRC d]es not perform this research there may be a need for a set of regulations requiring industry to deal with the uncertain aspects of component behavior under various fire con-ditions.

Sincerely, M e k David A. Ward Chairman 50

ft% os UNITED STATES d

8 3 NUCLEAR REGULATORY COMMISSION M  : c ADVISORY COMMITTEE ON REACTOR SAFEGUAnDS

1. W ASHINGTON, D. C. 20555 s...../ May 13, 1986 Mr. Victor Stello, Jr.

Executive Director for Operations

! U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON PROPOSED RESOLUTION OF USI A-17, " SYSTEMS INTERACTIONS IN NUCLEAR POWER PLANTS" During its 313th meeting, May 8-10, 1986, the Advisory Committee on Reactor Safeguards reviewed the draf t of a proposed resolution (Reference 1) for Unresolved Safety Issue A-17 (USI A-17), as presented to the CRGR on April 10, 1986. In its review, the Committee had the benefit of discussions with representatives of the NRC, as well as subcommittee meetings held on November 6, 1985, March 12 1986, and May 7, 1986. Furthennore, there has been a long history on this subject. The ACRS identified this issue formally in a letter dated November 8, 1974 to the then-Director of Regulation, L. M. Muntzing (Reference 2).

In brief, we believe that the proposal made to the CRGR on April 10, 1986 is inadequate. However, in view of the importance of the issue, we do not wish to defer all action on the resolution of this issue while a completely revised resolution is prepared over a period of many years. We recomend that the proposed resolution be modified to include a two-step process as indicated below and then published for public comment.

Recomendations

1. For the near-term, we recommend a resolution of the issue similar to that proposed in the August 13, 1985 draft CRGR package (Reference 3) sent from T. P. Speis to R. Bernero, J. P. Knight, H. L. Thompson, and W. T. Russell for comment. The NRC should clearly identify the limi-tations and scope of this near-term, partial resolution and state that the broader scope resolution will be developed over the next few years.

As proposed in the August 13, 1985 document, the near-term, partial resolution would include the following:

Some investigation would be required by licensees at all operating plants for certain specified spatially coupled adverse systems interactions.

• The NRC staff would provide data and background information on functionally coupled adverse systems interactions for information and possible application by licensees and/or industry groups, but with no immediate NRC requirement.

• The NRC staff would begin preparation of a new Standard Review Plan section on systems interactions to ensure a more integrated and 51

Mr. Victor Stello, Jr. May 13, 1986 systematic review of plants, initially for spatial, and later for functional, interactions.

2. For the longer term, the NRC should allocate the necessary resources and l initiate appropriate studies on functional and spatial interactions, l including causally connected failures, cascades, and other multiple I failure modes. This program of studies should be designed to ascertain, within a few years, which functional and spatial interactions are likely sources of significant risk. For these failure modes, a pragmatic ,

method of investigation and evaluation should be developed, tested, and then proposed, if appropriate.

For the longer term, consideration should also be given tc the use of a diverse and/or dedicated system for emergency heat removal which can serve as an acceptable resolution path for many aspects of the systems interactions issue while serving other safety functions.

3. We note that the proposed resolution of USI A-17 includes additional guidance for future plants in the area of single failure. If invoked, this appears to require a three-train design for those systems wherein a single initiating event can cause both a requirement for plant shutdown and a loss of one train of shutdown capability when, in addition, a credible single failure is assumed. We are inclined to favor the general idea. However, we recommend that such a far-reaching departure from past design practice be removed from the USI A-17 package and proposed on its own merit in a more appropriate forum.

Discussion

1. We conclude that the NRC staff has chosen a rather limited concept of component failure when analyzing for the possible effects of systems interactions. In the case of room flooding, for example, the effect on electrical components is defined in terms of component availability (or loss of availability) under the flooding conditions, and the resulting effect on safe shutdown capability. One is not required to consider l whether such flooding could result in electrical faults or spurious control signals that might interact through the electrical or control complex and thereby lead to a malfunctioning of equipment required for safe shutdown. We cannot support such a nonconservative bounding of the analysis.

We appreciate possible faulting that modesadequate (disruptions databeyond may not be available simple concerning)the loss of function of nonqualified components under flooding conditions, but the matter should be analyzed, using a range of plausible assumptions, instead of disre-garding the likelihood of faulting. We reconmend that the approach consider the faultina modes and consequential adverse systems inter-

actions associated with exposure of components to both safety and nonsafety conditions beyond their design basis, 52

Mr. Victor Stello, Jr. May 13, 1986 i

We recommend that particular attention be given to those cases wherein multiple adverse systems interactions are likely from a single comon cause event such as fire, pipe rupture, or earthquake. Such inter-actions may exacerbate an already hazardous situation and are likely to be numerous and appear in an unpredictable time sequence. We recognize l

i that such analysis could become very complex due to the numerous possi-ble interaction sequences, and that simplifying assumptions will be necessary.

Event-created environmental conditions beyond the component design basis have not been given adequate consideration in the proposed NRC staff resolution.

2. The proposed NRC staff resolution would fragment the systems inter-actions issue in its consideration by the NRC staff. There would probably be a similar fragmentation by licensees. This fragmentation has historically led to many systems interactions being neglected. We believe that the converse should be accomplished; namely, that both within the NRC staff and within each licensee's organization there should be an interdisciplinary group responsible for identifying and investigating systems interactions issues.

3. The proposed NRC staff resolution presented to the CRGR on April 10, 1986 would result only in an information letter identifying for util-ities a range of potentially significant areas for them to investigate and evaluate, relying largely on prior findings and some studies.

Although the concept of licensee responsibility for safety is at the heart of the NRC regulatory approach, in practice the record of licensee performance is uneven. There is little reason for confidence that, on the issue of systems interactions, all licensees will perform ade-quately. For this reason, among others, we support the proposed reso-lution of August 13, 1985.

4. We recognize that certain aspects of the systems interactions issue are currently treated in several sections of the Standard Review Plan.

However, the guidance for these matters is not complete. There are many gaps, and the USI A-17 Task Force itself concludes that many issues remain to be studied on existing plants.

Sincerely, s %

David A. Ward Chairman 53

Mr. Victor Stello, Jr. May 13, 1986  ;

References:

1. Memorandum dated March 3,1986 from Themis P. Speis, Director, Division of Safety Review and Oversight, for Raymond F. Fraley, ACRS,

Subject:

Draft CRGR Package on USI A-17

2. Letter from W. R. Stratton, Chairman, ACRS, to L. M. Muntzing, Director of Regulation,

Subject:

Systems Analysis of Engineered Safety Systems, dated November 8, 1974

3. Memorandum dated August 13, 1985 from Themis P. Speis, Director, Divi- '

sion of Safety Technology, for R. Bernero, J. P. Knight, H. L. Thompson, W. T. Russell,

Subject:

CRGR Package for Proposed Resolution of USI A-17 " Systems Interactions in Nuclear Power Plants" l

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UNITED STATES

!(fa atg[og n NUCLEAR REGULATORY COMMISSION N .E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON. O. C. 20555 o....

September 17. 1985 Honorable Lando W. Zech, Jr.

Chainnan U.S. Nuclear Regulatory Commission Washington, D.C. 20S55

SUBJECT:

ACRS COMMENTS ON THE RESOLUTION OF USI A-46, " SEISMIC QUALIFICATION OF EQUIPMENT IN OPERATING PLA*!TS" During its 316th meeting, August 7-9, 1986, and its 317th meeting. ds September 11-13, 1986, the Advisory Comittee on Reactor Safe A-46, reviewed the proposed resolution of Unresolved During Safetyour Issue (US

" Seismic Qualification of Equipment in Operating Plants."

review, we had the benefit of discussions with representatives of the Seismic Qualification Utilities Group (SQUG) and the NRC Staff, as well as the benefit of the documents referenced. The ACRS Subcomittee on Reliability Assurance met on this topic on August 5,1986 in Washington, D.C.

Over the past several years, the ACRS has had the benefit of numerous other briefings on the status of USI A-46. On each occasion we endorsed the approach being pursued by the NRC Staff in conjunction with SQUG as being appropriate.

USI A-46 applies to all nuclear plants not licensed under current seismic qualification practice as defined in Regulatory Guide 1.100, IEEE Standard 344-1975, and Section 3.10 of the Standard Review Plan.

Although we are still in agreement with the general approach for resolu-tion of USI A-46, we have the following concerns relating to the final resolution and the proposed implementation procedure. Where the con-cerns relate to components, nost of the items are nonseismically quali-fied components outside of containment.

1. The required seismic adequacy review will identify the minimura set of plant equipment required for safe, hot shutdown following an earthquake. This minimum set will be protected from seismically induced failure of other equipment which could threaten the integ- ,

rity or operability of the set. It is our understanding that the adequacy review will include a search for such failures, but will be limited to equipment whose failure can result in direct physical impact or electrical interaction with the set. Omitted are indi-rect interactions such as may be caused, for example, by flooding from the failure of a nonseismically qualified tank or pipe in the vicinity but beyond the physical impact range. We believe that a search for such indirect interactions should be included.

55

Honorable Lando W. Zech, Jr. September 17, 1986

2. When performing a seismic adequacy review, it appears logical to consider the seismic event as a simultaneous challenge to all plant structures, systems, and components. All credible accompanying failures should be considered as concurrent when forecasting consequences. It is not clear that such a requirement exists. It should be recognized that past studies have explored safety issues, such as flooding from pipe rupture, on the basis of only one such event occurring at a time. The safety conclusions from these studies will not necessarily apply for the seismic case wherein a combination of events may occur simultaneously.

3. The absence of any mention of seismically induced fires is notewor-thy. We believe this issue should be explored in sufficient depth, within the scope of the final resolution, to justify why such fires need not be considered, if this is indeed the case.

4. In addition, seismically induced actuations of fire protection features are not within the scope of the USI A-46 resolution. No reasonable justification is presented to support this omission. It may be inappropriate to rely on 10 CFR Part 50, Appendix R analy-ses, since they include an inadvertent actuation as a singular event for which safe shutdown may be achieved by redundant equip-ment located beyond the affected area. For the seismic case, multiple actuations may be possible and at least one inadvertent actuation might directly affect the minimum set of equipment required for safe shutdown. It appears to us that the kinds of components and control arrangements found in fire protection features could be susceptible to the failure modes identified by SQUG for other nonseismic equipment. Therefore, inadvertent actuation and its full impact on the equipment environments should be included in the required seismic adequacy review.

5. It is our understanding that the assumption is made that, irrespec-tive of size or seismic qualification, there will be no seismically induced failures of high-energy pipes. The proposed resolution remains silent on moderate energy pipes.) (We believe that this m be an unrealistic assumption for smaller, high- (or moderate) energy pipes unless supported by a plant-specific study which verifies that even the smallest air, instrument, or water lines will not fail, and are not susceptible to physical interaction damage resulting from failures of nearby nonqualified structures or components. If such failures can occur, they should be considered concurrent with the earthquake and any adverse environmental effects fully analyzed. If such failures result directly or in-  :

directly in a small LOCA, then the adequacy of the minimum set of I safe shutdown equipment to cope with its consequences should be included in the analysis.

56 1

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September 17, 1986 Honorable Lando W. Zech, Jr.

As of this date, the final resolution of USI A-46 has not yet been We presented to the Comittee to Review Generic Requirements (CRGR).

l wish to be informed of the outcome of the CRGR review, i

Additional coments by ACRS Member David Okrent are presented below.

Sincerely, 4

David A. Warc l

4 Chairman

! Additional Coments by ACRS Member David Okrent

~

It is not clear that the background of empirical experience for compon-ents beyond the first eight classes studied is sufficiently compre-I hensive to cover its application to other classes of equipment.

recomend that the Senior Seismic Review and Advisory Panel (SSRAP) be asked to provide detailed review and coments in writing for each class i

of components, spelling out the strengths and deficiencies of the data for the surrogate qualification task. Individual members of SSRAP j

i should be encouraged to identify any and all points of concern.

The Staff appears to be subsuming the issue of relay chatter into USI A-46 for relay perfonnance at the SSE level. I believe that this aspect I

of the work requires input from PRA studies into how serious relay chatter would be. Considerable assurance should be demonstrated that

relay chatter would not cause serious consequences.

l Many of the older plants, which have had the benefit of seismic walk-downs as part of USI A-45, exhibited significant deficiencies, including anchorages, battery supports or spacers, and some vital water tanks. I l

am concerned that there may still be a significant number of operating

plants which have such deficiencies and that they are not being un-l covered and remedied promptly. I recomend that the Staff minimize as

' much as possible the time by which all likely candidate plants have the i benefit of seismic walkdowns, without waiting until other aspects of

A-46, A-45, etc. have been resolved.

References:

1. U.S. Nuclear Regulatory Comission, " Seismic Qualification of

~

Equipment in Operating Nuclear Power Plants," Unresolved Safety Issue A-46, USNRC Report NUREG-1030 (Internal Review Version),

transmitted by memorandum from T. P. Speis, NRC, to NRC Division l

Directors dated June 30, 1986

2. U.S. Nuclear Regulatory Comission, " Regulatory Analysis for l

Proposed Resolution of Unresolved Safety Issue A-46, Seismic Quali-fication of Equipment in Operating Plants," USNRC Report NUREG-1211 57

. . - - . . _ . . _ , . _ _ , _ _ . , . _ _ _ . - _ _ - - - , ,,-. -,,_.-_,, - - ., - --~..-..-.m- -

l Honorable Lando W. Zech, Jr. September 17, 1986 (Internal Review Version), transmitted by memorandtn1 from T. P.

i Spels, NRC, to NRC Division Directors dated June 30, 1986

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>2 EIG UNITED STATES ,

j

[o 8 o NUCLEAR REGULATORY COMMISSION '

{e .E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS W ASHINGTON, D. C. 20555 September 17, 1986 l

Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON PROPOSED RESOLUTION OF GENERIC ISSUE 124,

" AUXILIARY FEEDWATER SYSTEM RELIABILITY"

During its 317th meeting, September 11-13, 1986, the Advisory Comittee on Reactor Safeguards met with representatives of the Office of Nuclear Reactor Regulation (NRR) to discuss a proposed resolution approach The for Generic Issue 124, " Auxiliary Feedwater System Reliability."

Comittee initially commented on this issue in a December 10, 1985 letter to W. J. Dircks EDO. Since that time, subcomittee meetings were held on this issue on March 26 and September 9, 1986.

)

As we noted in our December 10, 1985 letter, we believe the issue of j auxiliary feedwater (AFW) system reliability is important and deserves attention by the NRC Staff on a high-priority basis.

During the March 26, 1986 ACRS subcomittee meeting, NRR described a resolution approach that would require P acceptable AFW system reliability ( <10~g!R unavailability licensees on demand) as to dem stipulated in the Standard Review Plan (NUREG-0800). Demonstration of acceptable reliability would be shown by a probabilistic assessment, and continuing system reliability would be assured by a periodic (five-year) analysis of component failure data for the AFW system at a given plant.

Systems which could not be demonstrated to have suitable reliability would require modification to assure compliance.

The Office of Nuclear Reactor Regulation has now revised its resolution approach. As discussed with the Committee, the new approach involves an intensive review of the AFW system reliability for seven older operating l units. These units have two-train AFW systems which were judged to have questionable reliability, based on a review required after the TMI l accident. As we understand the new NRR approach, the reliabilityThe of each plant's AFW system will be evaluated by an NRC Review Team.

team will use information obtained from the Office of Inspections &

Enforcement inspections, plant visits, and other relevant data sources.

The owner of one of the plants under review, Prairie Island, has con-ducted an extensive reliability analysis of its AFW system. NRR indi-cated that the Prairie Island analysis will be a sort of benchmark against which the other plants will be measured, but no objective criteria for judging acceptability have been established.

59

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Mr. Victor Stello, Jr.

i September 17, 1986 t

.I l While the above studies will provide useful data, we are concerned with the lack of a definitive criterion for judging the acceptability of ArW i,

systems at these plants. Absent an objective benchmark or standard of i acceptability, we believe a plant-to-plant comparison of the relative

! AFW system reliabilities may not provide a satisfactory approach for resolving this issue. We recomend that the NRC Staff consider return-i ing to the review process it described during the March 1986 subcom-l mittee meeting.

Sincerely, f

! , h,  %

David A. Ward 4

Chairman 1

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UNITED STATES NUCLEAR REGULATORY COMMISSION l )

! .I n

ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

• W ASHINGTON, D. C. 20486 ,

o ,f,

• • • • • November 13, 1986 l

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! MEMORANDUM FOR: Victor Stello, Jr.

i Executive Director for Operations i

i

) FROM: R. F. Frale , -

' Executive Director ACRS l l I

SUBJECT:

ACRS COMMENTS ON THE PRIORITIZATION OF THE FOURTH GROUP OF GENERIC ISSUES j

i j

l During its 319th meeting, November 6-8, 1986, the ACRS reviewed the I l

adequacy of the proposed priority rankings for a group of Generic Issues j

identified in the attached Table 1, and its comments are contained in the following attachments.

l

• Attachment I lists those issues for which the ACRS agrees with the priority rankings proposed by the NRC Staff.

l ' Attachment 2 includes a list of issues for which the ACRS agrees j with the priority rankings proposed by the NRC Staff, but has comments.

!

• Attachraent 3 identifies the Generic Issue for which the ACRS j disagrees with the NRC Staff's proposed priority ranking along with the reasons therefor.

1

! Connents on Generic Issue 61, "SRV Line Greak Inside the BWR Wetwell Airspace of Mark I and Mark 11 Containments," have been deferred pending i j additional review by the ACRS. l

It is requested that the NRC Staff provide written responses to the ACRS  !

I comments identified in Attachments 2 and 3. t

) The ACRS will continue its review of the adequacy of the proposed i priority rankings for additional Generir Issues when they become

) available.

i Attachments: As Stated l

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! TABLE 1 1 1

GENERIC ISSUES REVIEWED BY THE ACRS DURING THE 319TH, NOVEMBER 6-8, 1986 MEETING f

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1 l GENERIC PRIORITY RANKINGS REFERENCE ,

! ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT L

l NUMBER 21 Vibration Qualification of DROP Memorandum from i Equipment Denton, dated i

{ June 23, 1986 61 SRV Line Break Inside the RESOLVED Memorandum from i BWR Wetwell Airspace of Denton, dated 1

Mark I and Mark !! Con- August 8, 1986  ;

tainments I 74 Reactor Coolant Activity DROP Memorandum from

! Limits for Operating Reactors Denton, dated j May 30, 1986 4

1 103 Design for Probable Maximum NEARLY RESOLVED Memorandum from i Precipitation Denton, dated i September 4,1985 111 Stress Corrosion Cracking of LICENSING ISSUE Memorandum from Pressure Boundary Ferritic Denton, dated Steels in Selected Environ- November 22, 1985 j ments i 114 Seismic-Induced Relay Chatter Covered in USI A-46, Memorandum from i " Seismic Qualifica- Denton, dated i tion of Equipment in June 25, 1986 Operating Plants" 115 Enhancement of the Reliability HIGH Memorandum from

of Westinghouse Solid State Denton, dated Protection System July 7, 1986

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+=-N+7-w T--=wW.*-ie-gas p.y9 tw+-wwweNN=1*WN-='N9 '-wW-ree**+vye--w-a-- p.-g-eT T*ir---s-i*g-,wg wey 9---,g.-,y---. -,-.4.-g-we,m._- 9.- , . . --iwee..gm-n--a. *

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] TABLE 1(Cont'd) i

! GENERIC PRIORITY RANKINGS REFERENCE ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT l NUMBER i

i 122 Davis-Besse loss of All Memorandum from j Feedwater Event - Short-Term Denton, dated j Actions January 28, 1986 1

l 122.1.a comon Mode Failure of HIGH

! solation Valves in Closed Position

! 122.1.b Recovery of Auxiliary Feed. MEDIUM l water

] 122.1.c Interruption of Auxiliary HIGH Feedwater Flow I

l 122.2 Initiating Feed-and-Bleed HIGH 122.3 Physical Security System LOW i Constraints

125 Davis-Besse loss of All Memorandum from Feedwater Event - Long-Term Denton, dated Actions June 30, 1986 125.1.2.a Need for a Test Program Covereu in Generic

! to Establish Reliability of Issue 70, "PORV and the PORV Block Valve Relia-bility"

! 125.I.2.b Need for PORV Surveillance Covered in Generic l Tests to Confirm Operational Issue 70, "PORV and j Readiness Block Valve Rella-

! bility" 125.I.2.c Need for Additional Protection RESOLVED

! Against PORY Failure f

l 125.I.2.d Capability of the PORV to Covered in US! A 45,

! Support Feed-and Bleed " Shutdown Decay Heat i

Removal Requirements" 1

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1 TABLE 1(Cont'd)

GENERIC PRIORITY RANKINGS REFERENCE l ISSUE TITLE PROPOSED BY THE STAFF DOCUMENT NUMBER l 125.11.3 Review Steam /Feedline Break DROP Memorandum from Mitigation Systems for Single (SafetyConcerns Denton, dated Failure of this issue have August 27, 1986 been addressed in Generic issues 125.II.1.b and 125.!!.7) 125.11.4 Thermal Stress of Once Through DROP Memorandum from Steam Generator Components Denton, dated Sept. 10, 1986 125.!!.7 Reevaluate Provisions to Auto- HIGH Memorandum from matica11y Isolato feedwater from Denton, dated Steam Generator During a Line Sept. 10, 1986 Break 125.!!.9 Enhanced Feed and-Bleed Covered in USl A-45, Memorandum from Capability " Shutdown Decay Heat Denton, dated Removal Requirements" August 27, 1986 125.11.14 Remote Operation of Equipment LOW Memorandum from Which Must Now be Operated Denton, dated Locally August 27, 1986 C-4 Statistical Methods for REGULATORY IMPACT Memorandum from ECCS Analysis ISSUE (RESOLVED) Denton, dated .

June 23, 1986 l

I C5 Decay Heat Update REGULATORY IMPACT Memorandum from ISSUE (RESOLVED) Denton, dated June 23, 1986 C6 LOCA Heat Sources REGULATORY IMPACT Memorandum from ISSUE (RESOLVED) Denton, dated June 24, 1986 l

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r ATTACHMENT 1

, LIST OF GENERIC ISSUES FOR WHICH l THE ACRS AGREES WITH THE PRIORITY RANKINGS PROPOSED BY THE NRC STAFF GENERIC ISSUE NO. TITLE 21 Vibration Qualification of Equipment 111 Stress Corrosion Cracking of Pressure Boundary Ferritic Steels in Selected Environments 122.1.a Comon Mode Failure of Isolation Valves in Closed Position l

122.1.b Recovery of Auxiliary Feedwater 122.1.c Interruption of Auxiliary Feedwater Flow 122.2 Initiating Feed-and-Bleed i 122.3 Physical Security System Constraints 125.I.2.a Need for a Test Program to Establish Reliablity of l

the PORV i 125.1.2.b Need for PORV Surveillance Tests to Confirm Operational Readiness 125.1.2.c Need for Additional Protection Against PORV Failure 125.1.2.d Capability of the PORV to support Feed-and-Bleed i

125.!!.3 Review Steam /Feedline Break Mitigation Systems for j Single Failure 125.!!.4 Thermal Stress of Once-Through Steam Generator i Components i

125.!!.7 Reevaluate Provisions to Automatically Isolate Feedwater from Steam Generator During a Line Break 125.11.9 Enhanced Feed-and-Bleed Capability 125.II.14 Remote Operation of Equipment Which Must Now be Operated Locally C-4 Statistical Methods for ECCS Analysis C-5 Occay Heat Update C6 LOCA Heat Sources 65 i

ATTAC MENT 2 LIST OF ISSUES FOR WHICH THE ACRS AGREES l WITH THE PROPO5ED PRIORITY RANKINGS, BUT WITH CL...JI3 j I

Generic Issue No: 74

Title:

Reactor Coolant Activity Limits for Operating Reactors l Priority Ranking Proposed By The NRC Staff: DROP ACRS Coments: The ACRS agrees with the proposed priority ranking for this issue. However, it offers the following coment:

Although the ACRS concurs, in general, in the NRC Staff's assessment of the expected savings in occupational radiation expo- ,

sures at Boiling Water Reactors (BWRs) I owing to the assumed implementation of more stringent controls, consideration might be given to refining these calcu-lations when data for BWRs, similar to those published in NUREG/CR-4485, "The Impact of Fuel Cladding Failure Events on Occupational Radiation Exposures at Nuclear Power Plants," for pressurized water reactors, become available.

66

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Generic j Issue No: 103

Title:

Design for Probable Maximum Precipitation

{

Priority Ranking j Proposed By The NRC Staff: Nearly Resolved ACRS Comments: Both the description and the resolution of j this issue are unclear.

The technical issue of what probable maximum precipitation (PMP) values should I be used to determine flood levels at i reactor sites would appear to be moot since the Staff has traditionally relied on the expertise and reconnendations of l NOAA for such values.

The regulatory issue of whether current NOAA values of PMP should be used to determine flood levels at existing plants has not been resolved. A decision has j been made by the Staff that a request for i new calculations of flood levels at NTOL

) plants is a backfit* and thus is not required; such calculations are being made

! by the Staff rather than being requested l of the applicant. Presumably, no such 1 calculations have been requested or are

! being made by the Staff for existing plants.

.I

, The ACRS believes that the safety issue of i possible flooding at plant sites has not

! been resolved for either existing or NTOL

! plants. Apparently, it has been resolved J for future plants b changes in the Standard Review Plan. y 1

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• The ACR$ has difficulty in understanding how the backfit rule can be applied because of the absence of any accepted means of quantifying

, the risk.

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ATTACHMENT 2 (Cont'd) i Generic Issue No: 115

Title:

Enhancement of the Reliability of Westinghouse Solid State Protection System Priority Ranking Proposed By The NRC Staff: HIGH ACRS Connents: The ACRS has concerns about some parts of the Staff's analysis in connection with this Generic Issue:

• Data used in the analysis appear to result from the treatment of testing-induced failures as random failures.

j Given this approach, if one chose to j increase the testing rate in an i effort to increase reliability, the failure rate would probably increase as well.

• The analysis of risk reduction appears to put principal emphasis on the operation (or lack of operation) of the undervoltage trip system.

Since experience indicates that this trip system is less reliable than the shunt trip system, the approach seems incomplete. The ACRS recognizes that the original intent of an undervolt-age trip system was to satisfy a

" fail safe" criterion. However, notable failures have occurred in this system that were not " fail safe."

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l ATTACHMENT 3 s

GENERIC ISSUE FOR WHICH THE ACRS DISAGREES WITH THE PRIORITY RANKING PROPOSED BY THE NRC STAFF Generic Issue 3: 114

Title:

Seismic-Induced Relay Chatter Priori ty ~ Ranking' Proposed by the x NRC Staff: , Covered in USI A-46, " Seismic Qualification of

' Equipment in Operating Plants" ACRS Comments: The Staff has concluded that the work being done in connection with resolution of USI A-46 and the Seismic Margins program as well as related programs cover the intent of Generic Issue 114 and therefore it need not be pursued as a separate issue. The ACRS does n.ot agree and finds the proposed resolu-tion to Generic Issue 114 unacceptable, for the following reasont:

Generic Issue 114-was to address the effects of seismic-induced relay chatter upon the safety and safety-related electrical and control systems as applied to all plants. USI A-46 o applies to and is limited to operating plants which were docketed prior to 1972. There is no indication that USI A-46 will be expanded to cover plants docketed since 1972. Therefore, the ACRS concludes that these plants will not have been adequately reviewed for the effects of seismic-induced relay chatter.

There is no indication that the Seismic Margins Program and related programs will adequately address seismic-induced relay chatter for earthquakes above the SSE.

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o UNITED STATES

~g 8 n NUCLEAR REGULATORY COMMISSION 3 ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHWGTON, D. C. 20556 Q

• e..*

August 12, 1986 Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS COMMENTS ON THE NRC POLICY STATEMENT ON FITNESS FOR DUTY OF NUCLEAR POWER PLANT PERSONNEL

Dear Mr. Zech:

During its 316th meeting, August 7-9, 1986, the Advisory Comittee on Reactor Safeguards reviewed the NRC Policy Statement on Fitness For Duty of Nuclear Power Plant Personnel. This matter was also discussed during a meeting of the ACRS Subcomittee on Human Factors on July 15, 1986. We had the benefit of the document referenced.

The Comittee last addressed this matter in the letter to Nunzio J.

Palladino, NRC Chairman, on August 9,1983 entitled, "ACRS Report on the Rule Concerning Fitness for Duty of Nuclear Power Plant Personnel . " In that letter, we noted that the proposed rule would not apply to NRC inspectors and said that such an exemption was inappropriate.

The current Comission Policy Statement on Fitness for Duty of Nuclear Power Plant Personnel continues to exempt NRC employees who have unescorted access to vital areas in nuclear power plants. We are informed that the NRC Staff is developing a fitness for duty program which would apply to these em-ployees. We applaud the NRC Staff initiative and recomend that the Comis-sion encourage continued development of the program and place it into effect as soon as possible. We wish to be kept informed of its progress.

Further, the Comittee endorses the random chemical testing of body fluids as an element in effective fitness for duty programs and encourages its incorpo-ration by nuclear utilities into nuclear power plant fitness for duty pro-grams.

Sincerely, O

t

.1 David A. Ward Chairman

Reference:

SECY-86-153, " Fitness For Duty of Nuclear Power Plant Personnel," memo for  ;

the Comissioners from V. Stello, Jr., Executive Director for Operations, l dated May 14, 1986 71

gwo uq'o UNITED STATES

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8 o NUCLEAR REGULATORY COMMISSION 3 ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 March 18, 1986 l

Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON DRAFT TECHNICAL REPORT ON GUIDELINES FOR BWR COOLANT PRESSURE BOUNDARY PIPING During its 311th meeting, March 13-15, 1986, the Advisory Comittee on Reactor Safeguards heard a report of its Subcomittee on Metal Compo-nents regarding the draft report NUREG-0313, Rev. 2, " Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping" (manuscript completed September 1985). We have no objection to the issuance of this report for public coment, and we offer the following three comments on the proposed guidelines.

1. We recommend that NUREG-0313, Rev. 2, be modified to limit the fraction of welds in a piping system that can contain cracks larger than acceptable under ASME Section XI IWA-3500 (with or without repairs). This fraction might be set at 25 percent. Cracks in primary piping represent a degradation of the system. Even though eacn repair is evaluated conservatively, their multiple interaction cannot be treated with any certainty. At some level of cracking, the piping should be replaced instead of patched again.

2. Welds made with IGSCC resistant materials can still crack in service, especially in systems without good water chemistry. Thus, we would recommend that weldments not be given " Category A" status (least frequent inspection) unless the IGSCC resistant material is combined with an approved stress improvement treatment or hydrogen water treatment.

3. Table 1 of NUREG-0313, Rev. 2, sets the maximum allowable carbon level for sensitization resistant 347NG stainless as 0.08 percent.

This should be reduced to 0.035 percent to aid weldability and l prevent the possibility of intense (" knife edge") corrosion at l welds.

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Mr. Victor Stello, Jr. March 18, 1986 Subsequent to the public comment period, we expect to review this report together with the public comments and the NRC Staff's response to them.

Sincerely, d

.l .

David A. Ward Chairman

Reference:

Office of Nuclea'r Reactor Regulation, Division of Engineering, "Techni- l cal Report on Material Selection and Processing Guidelines for BWR l Coolant Pressure Boundary Piping," Draft NUREG-0313, Rev.2, manuscript completed September 1985.

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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20S55 May 13, 1986 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON SALVAGING 0F CONTAMINATED SMELTED ALLOYS During its 313th meeting, May 8-10, 1986, the Advisory Committee on Reactor Safeguards reviewed the recent Comission actions relative to the Department of Energy (D0E) request for an exemption to permit salvaging of contaminated smelted alloys. We also reviewed the cuestions raised by Comissioner Bernthal on this matter. Supporting the Comittee in its deliberations were the discussions held by our Waste Management Subcommittee on April 24, 1986.

In our opinion, the question of the release of contaminated materials from enrichment plants is but a small part of a larger generic question concerning the disposition of a wide range of related materials.

Included in the latter would be metals and equipment resulting from the decontamination and decommissioning of nuclear power plants. For these reasons, and because of the lack of definitive data supporting the DOE application, we concur with and fully support the Commission in denying the DOE request.

We also concur with Comissioner Bernthal that a full range of alterna-tives for disposition of the alloys was not considered, especially those that could ensure definite barriers to the use of unacceptably contam-inated alloys in commerce. We fully support your plan to assist in the development of an integrated Federal policy which would establish consistent guidelines for dealing not only with the general issues associated with contaminated smelted alloys, but also with the issues associated with materials resulting from the decontamination and decom-missioning of nuclear facilities as well as the recycling and reuse of decontaminated lands, materials and equipment.

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Honorable Nunzio J. Palladino May 13, 1986

. We plan to be involved with and to report to you periodically as the work of the NRC Staff on these matters progresses.

Sincerely, ei b.

David A. Ward Chairman l

l

References:

I Memorandum for R. F. Fraley, Executive Director, ACRS, from R. B.

Minogue, Director, Office of NRR,

Subject:

Implementation of Commission Action: Denial of DOE Request for Exemption to Permit Salvaging Contaminated Smelted Alloys (with Attachments), dated March 17, 1986 l

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~g 8' n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

{o,  : ,E g WASHINGTON, D. C. 20555

September 17, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON DEGRADED PIPING RESEARCH During its 317th meeting, September 11-13, 1986, the Advisory Comittee on Reactor Safeguards heard a report from its Subcommittee on Metal Components concerning a recent meeting at which the research program on degraded piping was reviewed. Our coments are based on that report.

The research on piping integrity is related to the NRC's need for answers to the following questions concerning leak-before-break analy-ses:

- What flaw size can be accepted during operation?

- Should the postulation of a guillotine break in piping continue to be required?

- What materials and environmental control should be used in BWR piping?

The quality of the work was very good; thus, the connents that follow deal with how the programs fit in with NRC needs or might be improved.

Because the NRC Staff presentations rarely mentioned the objectives of-the programs in terms of NRC needs, the ultimate usefulness of the results was not made entirely clear, nor was it evident that these objectives are clearly understood by the people doing the work. It was uncertain how the necessary interaction with NRR is being handled in order to apply the results to the questions listed above.

The application of these results to piping systems in the field, as in Subsection IWB-3640 of the ASME Code, is predicated on the following assumptions:

- the crack is in straight pipe or a weld therein, I -

the calculations of P and P analysis)areconservEtive,b(fromthepipingsystem

- torsional moments do not affect allowable flaw size, and 1

77

Mr. Victor Stello, Jr. September 17, 1985 dynamically applied loads produce the same results as quasi-static.

The proposed International Piping Integrity Research Group (IPIRG) program will address the last of these, but in view of the uncertainties inherent in these assumptions there must remain some residual concern with the routine application of Subsection IWB-3640.

The program to which we would give the lowest priority is at Materials Engineering Associates, Inc. on " Fatigue Crack Growth Rate in PWR Materials and Environments." The questions being addressed concern the conservatism of the ASME Code Sections III and XI design rules and the fact that they ignore the effect of temperature and the water environ-ment. It is not clear whether the current approach leads to a conserva-tive or nonconservative design especially for plant life extension purposes. With.the demise of steel company research laboratories and the absence of new nuclear plant sales, the only group with interest and resources to pursue these questions seems to be the NRC and licensees.

To date, however, the tests needed to establish better criteria are in the process of being defined. The work on the mechanism of environ-mentally assisted cracking may be just developing, but the guiding ideas were not made clear.

Sincerely, f

. D.

David A. Ward Chairman 78

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UNITED STATES l 8 o NUCLEAR REGULATORY COMMISSION E ADVlsORY COMMITTEE ON REACTOR SAFEGUARDS

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August 12, 1986 Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS REPORT ON THE TENNESSEE VALLEY AUTHORITY'S MANAGEMENT REORGANIZATION AND SHUTDOWN OF TVA'S NUCLEAR POWER PLANTS During its 315th meeting, July 10-12, 1986, the Advisory Comittee on Reactor Safeguards met with representatives of the NRC Staff and Tennessee Valley Authority (TVA) and reviewed the issues related to reorganization of TVA's management structure and the shutdown of TVA's nuclear power plants. This matter was also discussed during the 316th ACRS meeting, August 7-9, 1986.

In this review, we had the benefit of a subcomittee meeting on June 12-13, 1986 and a visit to the Sequoyah Nuclear Plant on June 13, 1986. In ad-dition, we had the benefit of advice from two management consultants.

As a result of our discussions with the NRC Staff and TVA, it is our opinion that the immediate technical and management issues are being addressed and resolved. We agree with TVA's diagnosis that the root cause of their prob-lems was the lack of effective top management who could provide leadership and proper direction for TVA's nuclear activities.- There were a number of contributing factors which evolved over many years that ultimately created this problem, and it would not be surprising if corrective actions take many years to completely resolve this problem. We believe that the Comission should give this matter continuing attention.

In our discussions with TVA and the NRC Staff, we identified several issues that we believe must be dealt with if TVA's problems are to be corrected and their recurrence prevented. Our comments are as follows:

1. We were unable to identify the existence of an adequately structured

development program at TVA to assure that managers are prepared to handle the special problems associated with nuclear power plants and to provide for successful and systematic career development. TVA's Cor-porate Nuclear Performance Plan (Reference 1) does not describe or i express an intent to develop such a plan. TVA has taken some steps recently toward providing management development on an ad hoc basis and {

has for some time had training courses which could be used. We gen- '

erally agree that the measures taken by TVA with a temporary management

' team can satisfy their imediate needs while the contract managers are 1

79

Honorable Lando W. Zech, Jr. August 12, 1986 employed at TVA. However, we believe that a forward-looking, longer range, structured management development program needs to be established if TVA is to meet its future needs. The NRC Staff appears to share our Concerns.

2. In the present organization, the Manager of Nuclear Power has 25 indi-viduals reporting directly to him. This is clearly a large span of control and, if taken at face value, would be a serious deficiency in the organizational plan in the long term.

From our discussions with the TVA staff, it became apparent that there are shadow responsibilities amrt1 these organizational units. Some of the managers will actually rgort on most matters to the Manager of Nuclear Power through certain other managers. However, these lines of responsibility are not displayed in the organizational plan. We cannot tell whether they are clear to the managers involved. This arrangement may well be useful in the initial restructuring of the organization but will, in our opinion, need to be clarified if it is not changed in the process of transferring TVA's operations from the contract managers to permanent TVA employees. We recommend this matter to the Commission for their attention.

3. It is not clear where the focal point for nuclear safety resides within the TVA corporate management structure. The word " Safety" appears in several parts of the organization but the functional assignment seems fragmented with no clear assignment of responsibility. Although we agree that safety is everybody's business, we believe that a more focused approach is needed. It is not clear who has the responsibility for being the corporate nuclear " safety conscience" within the TVA higher management structure.

The former Nuclear Safety Review Staff appeared to have this mission at the TVA board level but did not perform it effectively. As a result, the organizational arrangement was discredited and the function has been absorbed by the Manager of Nuclear Power as a management review func-tion. We believe that TVA should reestablish credibility with regard to safety cognizance. The safety oversight process should be focused at a top management level. Such a measure may not be warranted for all licensees. TVA is, however, an unusual case and calls for unusual measures.

It is, in addition, not clear where the " safety conscience" of each site organization resides. The engineering groups seem to remain oriented toward design. The Independent Safety Evaluation Group (ISEG), while ostensibly filling the function of safety conscience, appears only to provide safety-oriented services. The group which is to play a key role in safety-related decisions and evaluations needs to be identified.

4. TVA's compensation program for top level nuclear managers is not compet-itive with the rest of the nuclear industry. This problem was recog-80

Honorable Lando W. Zech, Jr. August 12, 1986 nized by the President's Private Sector Survey on Cost Control, fre-quently referred to as the Grace Comission Report (see Reference 2),

which recomended that corrections be made imediately in order to hold together the nuclear program management team at TVA. TVA is, as recent experience shows, not able to compete in hiring in the market place except under special temporary contract arrangements. Until the statu-tory limitation is changed to permit TVA to pay salaries competitive with the industry, it will be difficult, even with development programs, to meet TVA's needs for experienced top level nuclear managers. We recommend that the Commission give attention to this matter.

5. TVA does not use modern personnel selection techniques like those used in other segments of the nuclear utility industry and many successful corporations. We recomend that TVA review the use of personnel se-1ection methods and utilize the most effective aptitude testing avail-able to them for personnel selection, transfer, and promotion.

6. The NRC Staff is currently confronting the difficult task of reviewing a large organization with complex and long-standing organizational prob-lems, and judging if adequate reorganizational steps have been taken.

The NRC Staff in the past has been oriented to hardware rather than management problems and perhaps has not given enough attention to developing a capability to perform management reviews. We believe that improving the management of activities related to nuclear power plants continues to be one of the most effective means for improving the safety of operation. We recomend that the Comission give high priority to developing NRC capability for conducting management reviews. Emphasis should be placed on identifying management problems before they lead to difficulties on the scale encountered at TVA.

7. In our discussions with TVA and the NRC Staff, we attempted to gain some perspective on problems associated with the restart of the TVA plants.

The items delaying the restart of Sequoyah appear to be resolvable in the near future. Little was presented during these discussions concern-ing the additinnal work required to prepare Browns Ferry for restart or for licensing of Watts Bar. A concerted effort should be applied by the TVA and NRC organizations to complete the necessary documentation and review of safety questions for Sequoyah and return those units to power at the earliest date that nuclear safety can be assured. Restoring power operation at Sequoyah should significantly improve the morale of the entire TVA nuclear organization and restore confidence in present management. We believe that high morale of nuclear plant operating personnel is another of the most important ingredients for plant safety.

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Honorable Lando W. Zech, Jr. August 12, 1986 Additional coments by ACRS Members Glenn A. Reed and David A. Ward are. '

presented below.

Sincerely, r

i

.i David A. Ward I Chairman Additional Coments by ACRS Member Glenn A. Reed Although I agree generally with the conclusions and recomendations of this

. ACRS letter, I am concerned that the ACRS has stepped out of its regular t

fields of expertise into the utility management field where it does not have a wealth of experience. The use of consultants by the ACRS was appropriate to the effort; but, even so, more in-house experience to evaluate the consul-tants' input could have been important to final judgments.

4 Consistent with my concern about the ACRS experience and training to review 1 and recomend on management issues, I disagree with the ACRS recomendation to have the NRC Staff strive to develop expertise to evaluate nuclear utility management. I consider better and more timely abilities already lie with or can be acquired by INP0. INP0 has been addressing this issue for some time.

Governmental agencies are not noted for their ability to lead the way by exemplary management structure and performance, and I doubt that even many

years of training and development of NRC personnel would bring the regulatory organization into a sound position to render advice and direction.. In my i opinion, the technically based NRC organization should continue to concen- l trate on performance indicators, and from these draw conclusions that there j are or are not management problems needing correction, and then leave the direction of the corrections to industry.

In this regard, I consider that the NRC regulatory organization (Comis-i sioners, headquarters staff, and regional staff) did a respectable job in focusing attention upon TVA performance problems in its nuclear operations.

i I consider that, in spite of much criticism to the contrary, the NRC did move l on a reasonably respectable schedule to indict and curtail TVA construction and operations activities.

Additional Coments by ACRS Member David A. Ward l Some of the problems experienced in the operation of TVA's nuclear plants have been attributed to the past dominance within TVA of the " architect-i engineer" perspective over the " operations" perspective. The reorganiza-tional approach and many of the temporary managers appear to be influenced by this same " architect-engineer" tradition. I have some concern about whether the " operations" point of view will be given an adequate voice in the new organization.

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Honorable Lando W. Zech, Jr. August 12. 1986

References:

1 1. Tennessee Valley Authority, Revised Corporate " Nuclear Perfomance Plan," dated March 1986 i

2. President's Private Sector Survey on Cost Control-Report on Boards /

Commissions transmitted to Hon. Ronald Reagan from J. Peter Grace, Chairman, Executive Comittee on September 15, 1983 4

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'o UNITED STATES d ~

8 o NUCLEAR REGULATORY COMMISSION 9 ,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ WASHINGTON. D. C. 20555

• • • • • March 20, 1986 Honorable Nunzio J. Palladino l Chairman U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

RESPONSES TO RECOMMENDATIONS OF PANEL ON ACRS EFFECTIVENESS

Background

During the latter part of 1984, the Advisory Comittee on Reactor Safeguards decided to undertake a concerted review of its own effectiveness as an advisory body to the Comission. We reflected on the role ACRS has.taken in

the past and considered whether continuation of that role, substantial modification, or even dissolution of the Comittee might be most appropriate for the future. We recognized that the Comission was at a turning point in its history. New plant design, construction, and licensing activities have

become less important; or the time to effectively influence them has passed.

Concern about operations in almost 100 nuclear power plants has become paramount. And this has proven to be something with which the body of regulations and reservoir of expertise within the agency is it.ss able to cope. Also, the need has arrived for regulations to deal with the largely political and social issues related to disposal of high-level waste. And finally, the proper plan and schedule on which the agency must prepare for a resumption in building of nuclear power plants, whether more of the same, standardized plants, or non-LWRs is very unclear and difficult to bring into focus.

> Given what seemed to us to be a different mix of challenges facing the Commission than that of ten years ago, we questioned whether the makeup of the ACRS, its methods for developing advice, and the subject matter it had traditionally addressed were still appropriate for effectively assisting the Comission with the essential issues of today. The Comittee discussed and agreed upon several changes in its operation and new initiatives in its programs. However, we recognized that a critique of our advisory role and its effectiveness from an outside perspective would be useful.

< Formation of Panel on ACRS Effectiveness Accordingly, we set up the " Panel on ACRS Effectiveness" and were fortunate in assembling a group of experienced and distinguished individuals. Eight of the nine panel members represented a spectrum of viewpoints and experience in i

nuclear safety, with each having an interest in improving reactor safety and 85 f

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Honorable Nunzio J. Palladino March 20, 1986 safety regulation. The ninth [Dr. Hagedorn] is a senior consultant in management and organizational effectiveness. The Panel membership consisted of:

Mr. L. Manning Muntzing, Chairman Dr. John F. Ahearne Mr. Myer Bender Mr. Edson G. Case Dr. Homer J. Hagedorn Mr. Richard Hubbard Dr. Herbert J.C. Kouts Mr. Steven Sholly Mr. John M. West The Panel was asked "to provide advice and guidance that will permit the ACRS to improve its effectiveness as an advisory body to the Nuclear Regulatory Commission and in furnishing more general leadership in matters of reactor safety." A fuller statement of the Panel's charter is given in Appendix A.

The Panel began its work in March of 1985 and presented its report to the Committee at our 305th meeting on September 12, 1985. Copies of their report have been furnished to you.

To supplement the prior knowledge of its members, the Panel gathered infor-mation about regulatory needs, ACRS programs, and ACRS procedures from systematic interviews of people concerned with and experienced in nuclear power, safety, and regulation. A total of about 65 interviews was conducted with present and past members of the ACRS and of the Commission, present and former NRC Staff, hearing board members, Congressional staff, representatives of industry and public interest groups, and State government authorities.

The Panel began its activities by developing an estimated scenario for the course of nuclear power and safety regulation through the year 2000 so that the role of ACRS could be considered against what needs to be done in the present and near future rather than what has been done in the past. The Panel's report is in several sections:

Scope of ACRS Activities Through the Year 2000 ACRS Approach to Substantive Technical and Related Issues Relationships of the ACRS Within and Outside the NRC i

Membership on the ACRS and the Selection Process Organizational Effectiveness of the ACRS ACRS Operational Methods Enhancing ACRS Contributions to the Regulatory Process Conclusions and Recomendations General Recomendations of the Panel The Panel recomended that the ACRS continue in its present role with some changes in emphasis and in its methods of operation, 86

March 20, 1986 Honorable Nunzio J. Palladino It should consider its primary task as advisory to theItCommission rather than.

should be prepared to to the public, the Congress, or the NRC Staff.

assist the Commission in the full range of technical issues with which the Comission is concerned. With regard to such issues, it should concentrate its advice on matters of broad policy and should not attempt to develop detailed solutions to problems.

The Panel also recommended that the ACRS continue to consider the development However, of consensus positions on issues as its primary mode of operation.

the Panel believed that ACRS reports should furnish more explanation of the reasons for positions the Comittee has taken and of the range of opinions which underlie the consensus.

The Panel further recommended that the ACRS retain its present size, but It recomended that the consider its composition against future needs.

Comittee strengthen internal leadership and that it consider whether the Comittee staff could be reduced given an emphasis on broad policy rather than on details of issues.

General Responses of the ACRS in Since receiving the Panel's report and recommendations, the ACRS has, several meetings of its Subcomittee on Procedures and Administration and of the full Comittee, reviewed and discussed them at length. We concur with the thrust of the Panel's report and conclusions and plan to adopt the overall sense of its recommendations.

Although we We agree that our primary mission is to advise the Comission.

have a specific obligation to advise the Congress on the research program, we perceive this to be a less demanding task than in the past and believe it does not seriously detract from our primary responsibility. We agree that our interests should be as broad as those of the Comission, but believe our resources should be concentrated on the issues which have greatest importance to public safety. We also note that while our advice should concentrate on broad policy, it is frequently necessary for the Comittee to delve into details to develop the proper understanding of important issues.  :

The Comittee has recently restructured its system of subcomittees to eliminate unnecessary project subcommittees and to provide increased emphasis 4

on operating plants and on issues of radioactive waste disposal. In addi-tion, we have created subcommittees and Comittee programs to advise the Comission on development of a long-range plan for the agency and to provide a statement on the state of overall safety in the nuclear power plants of j today.

The ACRS views its collegial nature and its ability to develop consensus positions as uniquely valuable attributes. While we agree that more ef-fective comunication of the reasons for our positions and of the opinions of 87 l

Honorable Nunzio J. Palladino March 20, 1986 individual members could be helpful to the Comission on occasion, we believe too much emphasis on the latter will seriously undennine the effectiveness of the ACRS.

The Comittee is cognizant of the need to add members with the appropriate background for dealing with current and expected future issues. However, we believe each member must also be able to function effectively as a general-ist. There are, and will continue to be, many more than 15 technical areas of concern with which the ACRS must deal. The Comittee has taken steps, formation of a management comittee and a trial two-year tenn for officers, to strengthen its internal leadership.

The Committee is accommodating to budget-inspired reductions in its staff by reducing support activities for members and by eliminating programs which are perceived by the Committee to be less important to public safety. We believe there is no reason to expect, as the Panel seemed to, that a reduction in staff would follow directly from an ACRS move toward concentration on " broad l

policy issues." Given any further staff and resource reductions, we will eliminate less important programs.

l The Panel made twenty-seven specific recomendations in three areas:

A. ACRS Mission B. Relations with the Commission and the NRC Staff C. Internal Operational Questions We find most of these specific recommendations to be wise and helpful and plan to adopt them, in spirit if not always to the letter. Powever, we take exception to a few. In most cases, the Panel's recomendation is to the ACRS and response is under control of the Comittee. In three instances (A.2, A.3, and B.3 in Appendix B), action by the Commission or the Congress would be required to fully comply with the recommendation. While we disagree with none of these three recommendations, we believe there is no urgency connected with any of them, and we plan to initiate no action in the near term. In Ap-pendix B we have repeated each of the specific recommendations and provided a sumary of the Committee response including a description of action planned.

Sincerely, i b.

David A. Ward Chairman Attachments:

Appendix A - Proposed Task Description for an ACRS Effectiveness Study Appendix B - Specific Panel Recommendations and ACRS Responses

. 88 1

' APPENDIX A i

Proposed Task Description for an j i ACRS Effectiveness Study l Background l In a meeting of its Subcommittee on Comittee Activities at Harpers Ferry in November 1984 and in the full Comittee meeting of December

1984, the Advisory Committee on Reactor Safeguards reviewed its own  ;

! performance and effectiveness and began consideration of several efforts  :

i to better define its role, goals, and procedures as part of a program to '

improve its effectiveness as an advisory body, including any statutory changes that may be required. The Committee has been active for over 25

years and during that time has witnessed great change in the nuclear l power industry and in the AEC/NRC regulatory role. While the ACRS has also evolved in its safety advisory role through the years, further change may be in order. Although a number of improvements in ACRS procedures and priorities have been agreed upon by the members, it also became apparent that it would be useful to engage an outside panel of experts to make an ad hoc appraisal of the present and anticipated future effectiveness of the ACRS and to develop recomendations for improving the' Comittee's performance.

Organizational Appr'oach

~

The A.CRS will appoint a panel of three to five persons who have back-

. ground and interest suitable for assessing the role of ACRS in today's a environme'nt. The Panel will be reimbursed, making use of NRC contrac-  ;

tural arrangements as

• appropriate. The Panel will have one of its members designated as Chairman, and he or she, in conjunction with the ACRS Chairman and the ACRS Executive Director, will formulate a general 4 charter for the review. The panel will arrange for its own administra-tive and secretarial support.

i l Schedule The Panel shouTd complete its review on a schedule that will permit it to make a report to the Comittee no later than July 1985.

Panel Membership:

Panel Chainnan - L. Manning Muntzing, Doub & Muntzing Law Fim i

John F. Ahearne, Resources-for the F'uture Herbert J. C. Kouts, Brookhaven National Laboratory Myer Bender, Querytech, Inc.

1 Homer J. Hagedorn - A. D. Little i

EdsonG. Case,NRC(retired)

JohnM. West,CombustionEngineering(retired)

Steven Shelly - Union of Concerned Scientists Richard Hubbard - MHB Technical Associates A1 l

i I 89 APPENDIX A  :

I J

4 Task

Description:

The purpose of the Panel's review should be to provide advice and guidance that will permit the ACRS to improve its effectiveness as an advisory body to the Nuclear Regulatory Comission and in furnishing more general leadership, in matters of reactor safety. There.are indications that some of the Committee's practices and policies, while effective in the past, may not be as appropriate for the maturing and changing industry and regulatory' climate of today. 1 A base for the review should be the Comittee's charter as described in the Atomic Energy Act of 1954, as amended. If the Panel believes this fundamental charter should be changed, recomendations for that should be clearly separated from advice as to how the Committee can more effectively fulfill its present charter.

Given this starting point, the Panel should consider itself unrestricted j in its review. We have provided below a list of questions that the 4

l Panel might explore. We have no praconception about the completeness or the essential nature of these questions; they are examples only. We expect that the Panel may want to r,eview recent and past ACRS reports, interview members of the Comittee, past and present, interview Comis-stoners and Comission staff, members of Congress and staff, and rep-resentatives of the nuclear power industry and the broader reactor safety comunity.

1. Is the Comittee's approach to generating advice on issues (i.e.,

through review of related documents supplemented by testimony to Subcomittees and the full Comittee, and then developing a sum-mary, consensus report to the Comissioners) appropriate and effective?

2. Is the Comittee's present composition and the method of member selection appropriate and effective?

3. Is the method of selecting the Comittee chairman and subcommittee chairmen and members the best approach?

1 4. Are follow-up activities, vigorous enough?

5. Should the ACRS work on generic vs. case-by-case determinations?

6. How should/should the ACRS interface 0with the NRC Staff,

'Comissioners, and boards be changed, if at all?

7. How snould the ACRS establish its work priorities?

8. Is the scope of Comittee involvement too broad or too narrow?

l A2 90

Should the

! 9. Does the ACRS make effective use of its consultants?

use of consultants be expanded / reduced?

10. Are the Committees conflict of interest restrictions too tight?

See guidelines being prepared by OGC and ACRS regarding financial and apparent conflict of interests (per telephone conversation with D. Ward on February 11,1985).

11. Should the ACRS devote an increased amount of its effort to following the activities of operating reactors and more direct / active involvement in the investigation, evaluation of reactor incidents and accidents?

12. Should operating experience gained from operating facilities be used as the basis for comment regarding the specific facility and/or more generic considerations of the program in general.

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APPENDIX B SPECIFIC PANEL RECOMMENDATIONS AND ACRS RESPONSES Recomendation A.I. "We believe that an advisory group such as the ACRS will still be strongly needed in the future. However, ACRS should turn its attention to concentrate on the broad technical policy questions 1 on which the Comission needs wise counsel. A few examples of i the kinds of issues that should be taken up by the Comittee are given in Chapter III of our report."

Response

We agree, with coment.

While we concur that the " output" [ advice] of the ACRS should be confined to matters of broad policy, we agree that we have not always limited ourselves thus in the past, and resolve to do better. However, we note that much of the " input" [information gathering] to the ACRS must be at rather detailed technical levels. We must maintain an effective sampling process among the trees to understand and appreciate the forest.

Recommendation A.2. "The ACRS should be relieved of the requirement to conduct reviews and issue reports on all nuclear plant applications. The Comittee should only have to review a license application when a safety-related feature is proposed that is new and significant.

In such cases, review should be necessary only for that feature."

Response

Requires Congressional action. We agree, but consider it not urgent. l We believe it is not an issue at present because of the rela-tively few applications pending. If there is a resumption in building and licensing plants, individual ACRS reviews may or may not be appropriate depending upon the nature of the resurgence.

We believe the Comission and the Congress should consider a change in the law for the future, but there is no present ur-gency. We plan no action on the part of ACRS to initiate a ,

change.

l 92

Recomendation A.3. "ACRS reviews of the safety research program should focus on the question of whether the right research programs are planned or underway, in light of priorities. The detailed budgetary-level review seems to serve no strong purpose, and should be dropped."

Response '

Requires Congressional action. We agree, but consider it not urgent.

However, we believe it is possible for the ACRS to prepare an annual report to the Congress, simpler and less demanding of ACRS resources than those of the past several years. We do plan to

' give continued attention to the research program. Our report to the Congress can be a spinoff from this with less emphasis on 7

details of the research budget and more on the general content and direction of the program. Therefore, unless we hear from the Congress that a more comprehensive annual report is desired, we will continue to honor the present mandate as described.

l Recommendation l A.4. "The scope of the Comittee's technical interest and attention should be essentially as broad as that of the Comission. The Commission should feel free to seek the Comittee's views on any technical regulatory problem confronting it."

Response

We agree, with coment and an exception.

We agree with the general thrust of the recommendation and believe the Comittee does and should continue to concern itself-with any issue identified by the Comission, or by the Comittee itself, which is important to nuclear safety. We do not agree 4 with the Panel's implication that the Comission should seek the

! Comittee's views only on " technical regulatory problem [s]."

j While what that phrase means might be debated, we believe a

likely connotation is too narrow. Historically, the Comission

has sought and heard Comittee advice concerning a broader range

, of issues. The tests of pertinence we intend to use are [1]

whether the issue is important to safety and [2] whether the

Comission will find the advice useful.

Recommendation A.S. "The ACRS will be most valuable in its revised role if it confines its recommendations to the directions to be taken, and avoids developing detailed solutions to problems. Its advice should be at the level of the forest, not the trees."

93

i F

Response

l We agree, with coment.

I We agree the ACRS should not attempt to develop detailed solu-

tions to problems but should confine its advice to general j directions and policy. However, we repeat the coment above. In 1

order for us to understand the forest and to make general recom-mendations about fores try', it is necessary that we sometimes immerse ourselves in the trees.

Recommendation i

J i A.6. "There are other roles the ACRS should not try to fill: i

) a. It should not try to respond to day-to-day questions. That j is, it should not do fire fighting."

i j Response We agree, f Recommedation

)

i "b. It should not try to be a kind -of National Transportation i Safety Board. If it reviews findings of investigative groups, it should do so for their technical policy

implications."

Response

i We agree.

! Recommendation "c. It is not a second Regulatory Staff, and should not try to function as one."

Response

We agree.

l Recommendation i

"d. It should not try to manage or oversee the NRC Staff."

Response

i We agree, with comment.

i

.i i 94 4

i

i There may be occasions on which we will coment on how well some part of the Staff is doing in dealing with some particular issue.

Recomendation i

"e. The special relationship it once had to Congress has ended, and no effort should be made to reestablish it."

Response

We agree.

Recommendation B.I. "The ACRS should be advisory primarily to the Commission, provid-l irig advice on technical policy as requested, and also gn the initiative of the Comittee." ,

Resporte l

We agree with coment.

We note that some of the beneficial influence of the Committee in the past has been through direct interactions with the Staff. We believe these are useful and should continue. We also note that our relationships with the part of the Staff reporting to the EDO, with whom we have a Memorandum of Understanding, have been more fruitful than relationships with Comission-level offices.

We intend to develop better working relationships, perhaps .

including MOVs, with some of these other offices.

) Recomendation

B.2. "The ACRS could also provide advice to the NkC Staff, but when this is done it should be through a communication addressed specifically to the Staff. Such advice may have the greatest value if it is provided early in the process of consideration of an issue."

Response

We agree, with coment.

We agree that advice to the Staff, particularly coments on pre-

liminary positions of the Staff revealed as part of Comittee-1 Staff discussions, should be addressed to the Staff and not to l the Comission. We intend to address such advice directly to the l l

Staff in the future, for example, by letter addressed to the EDO.

i 95

Recommendation B.3. "The Commission should consider, now and on a continuing basis, how it can better take advantage of the existence of an ACRS which is redirected to providing broad advice on technical policy questions."

Response

This is a recommendation directed to the Commission; however, we have a comment.

There are two potential ways in which the Commissioners might be tempted to use the ACRS, but which we believe would be a misuse of the uniqueness of the ACRS.

One would be to expect the Committee to respond to the fire-of-the-week that challenges the Commission and Staff. The ACRS, consisting of part-time members, sited remote from Washington and remote from each other is not well-suited to this sort of ac-tivity. The ACRS functions best in a more deliberative, hope-fully thoughtful, mode. The second is to use the experience and expertise of ACRS members as individual consultants. We perceive that the most unique capability of the ACRS is its inclination and ability to develop collegial positions on complex matters.

We believe that too-frequent individual advice and pronouncements from individual members to Commissioners or Staff will seriously erode the efforts of the Committee to reach consensus. There are many other individual consultants available to the Commissioners or the Commission.

We believe neither of these have been serious problems in the past, but we see trends indicating they might become such in the future. We encourage the Commission not to nuture these trends.

Recommendation B.4. " Oral communications between the ACRS and the Commission should be improved by better management of the agenda. Not all members of the ACRS need to attend every meeting with the Commission. A smaller group of those principally concerned with the subject to be discussed would enable better exchange and discussion."

Response

We agree only in part.

We believe that joint meetings between the Committee and the Commission have been useful and believe they should continue, perhaps with an increased frequency. liowever, we believe it is not desirable to limit attendance of ACRS members to fewer than the full Committee. We have seen no evidence that discussion has been impeded by the numbers. We think that the general practice 96

of full attendance by ACRS is an important protection of our collegiality.

We note that the impending relocation of Comission and ACRS offices to different suburbs is likely to discourage future joint meetings.

I i

Recomendation i B.S. " Individual Comissioners should be able to discuss ACRS matters with individual ACRS members, not to undercut the collegial i

nature of either the Comittee or the Comission, but to open a channel that would pennit the Comission to appreciate ranges of ,

opinion of ACRS members in rendering collegial advice to the t Commission. We see the pitfalls in this arrangement, but we believe they can be avoided. In the next section on internal operational questions, we recomend development of procedures for this one-on-one interaction."

Response

We agree, with coment.

I The ACRS is considering an addition to its bylaws to govern such interchanges. We are opposed to practices which amount to having i

an individual ACRS member " lobby" the Commission or a Commission-er on a particular issue.

Recomendation l

i B.6. " Relations between the ACRS and the NRC Staff should be improved.

Presumably a trend toward improved interpersonal relations can develop if the ACRS is redirected so as no longer to seem to be in competition with the Staff on licensing issues, or in the position of second-guessing the Staff."

Response i 1

We agree, with coment.

We acknowledge there have been occasions on which the ACRS i members have appeared to be rude to members of the Staff in the l course of, sometimes, heated discussions in public meetings. We ,

i also acknowledge that we are too sparing in articulation of

• 1 praise for the Staff even when it is obviously appropriate. We attribute this to our own huan fraility and inattention rather 2

than to any systematic fault it: the NRC Staff. To the contrary, 1 we believe the NRC Staff is, in the main, competent and profes-i sional and serves the nation well. 1 2 I

) 97

Recommendation C.I. "The Panel recomends that the role of the ACRS Chairman should be strengthened, and that the Chairman should exercise a stronger leadership role. As part of this strengthening process, the Chairman's term of office should be lengthened to two years, ,

renewable once, and he should receive a higher remuneration for his service."

Response

We agree, with an exception.

The ACRS agrees to a two-year trial period which has begun without a change in its bylaws. We will consider a permanent change after this trial. We disagree that higher pay for the Chairman is appropriate.

Recomendation C.2. "The ACRS operates at present in a somewhat reactive mode. The subcomittees on generic questions provide some long-term coher-ence to Comittee action, but a Management Comittee is needed to plan out future directions of the Comittee, develop priorities for issues to be considered, and set the course of action and schedule for these. A suitable Comittee might be the Chairman, the Vice-Chairman, and the past Chairman, with the Executive Director of the ACRS Staff as an ex officio member."

Response

s We agree, with coment.

The ACRS has formed a Management Comittee which functions' as the Panel recommended. However, we may alter the membership of the I

Management Committee somewhat from that recomended to' assure that no single person is a member for too long.

l Recommendation C.3. "The consensus method of operation is appropriate. However, dissent should not be discouraged. Papering over disagreement ,

with words meaning different things to different Comittee .

members does not help the Comission and does not strengthen the I image of the Comittee. Letters might well reflect the breadth of opinion of the Comittee. They should especially point up the opinions of any Comittee members who are particularly expert in

, matters on which the Comittee is comenting."

1 98 l l

Response

We agree, with coment.

The ACRS agrees with the Panel recomendation. We believe con-sensus is the most desirable mode when it does not result in watered-down or ambiguous advice. We have and use a method for dissent in our reports - separate coments by individuals. We believe this is adequate to assure dissent is heard. We plan to consider whether there can be some way to comunicate the range 1 of opinion, as opposed to actual dissent, contributing to the consensus expressed in our reports.

l Recomendation C.4. i'Comittee letters should be clear and unambiguous. Letters should be self-contained, so that readers need not refer back to previous letters for interpretation or understanding. If the Comittee moves away from responding to licensing needs, as we i suggest, letter writing can then be more thoughtful and the 4

product more meaningful. Subcomittee chairmen should be in a i position to explain the entire contents of a letter and not just a limited part."

Response

We agree, with an exception.

The Comittee is considering the development of guidelines to assist members who draft Comittee letters [usually the chaiman of a subcomittee which has developed the issue for the Comit-tee]. We do not believe it is always reasonable or necessary to i expect the subcommittee chairman to be able to answer all .

questions related to a given letter.

~

Once issued, a letter  ;

" belongs" to the full Comittee, not to the original drafter or

{

the subcomittee.

Recommendation

, C.5. "We see no reason to change the size of the ACRS. A membership of fifteen still seems reasonable as does the method of selection of members."

Response

We agree, with coment.

The ACRS accepts the recomendation. While 15 is a rather large committee by some theories of organizational effectiveness, the ACRS has enough tradition and experience that it is able to func-tion and reach consensus adequately. Too many fewer than 15 4

would make it difficult for the ACRS to deal with the wide range of issues it must accomodate.

99 4

,, - - + - - - - . - - - - -. -,- .-

s 1

Recommendation C.6. "The present policy that limits ACRS service to three tenns is reasonable. The practice of long-term tenure for Comittee members prevents taking on new members with fresh ideas or special competence in new areas of Comission interest."

i Response We agree, with comment.

We agree as a general practice. We may propose occasional and specific exceptions to the general rule if we believe there is justification. We also may propose greater use of emeritus status.

Recomendation C.7. " Reappointment of members should not be automatic. When a mem-ber's term ends, his reappointment should be considered in the same way and on the same plane as appointment of another person-to replace him."

, s

Response

We agree, with an exception.

We must acknowledge, however, that it is possible to judge a sitting member in light of his or her known contributions, and this may influence choices. There may sometimes be a tendency to prefer "the devil we know over the devil we don't know." In

> addition, when recomending a present member for reaopaintment, the ACRS does not intend to include the names of one or two other candidates as is the practice for first-time nominations.

ts Recomendation .

C.8. "The redirection of ACRS away from ' nuts and bolts' questions should reduce the need for the Comittee to have members repre-senting a complete range of disciplines. Some range of special-ist backgrounds may still be useful, but it will also be im-portant that Comittee members be generalists who can consider questions as broad issues. Thus, the ACRS should move away from requiring new members to be clones of the departing ones." j l

t I t 4

.( l 1

l s 100

k i

i

! l

Response

I We agree, with coment.

We agree that wisdom should be considered as of primary im-portance over detailed expertise; a specific discipline can be provided by consultants. Our intention is to consider candidates for nembership for their ability to contribute to the general work of the ACRS, as well as for their specific experience and expertise.

Recomendation C.9. "Though the amount of time devoted to ACRS business by individual Comittee members is in many cases very high, it should be made clear to prospective members that substantially less time could be adequate. The time comitment should not be a serious ob-stacle to recruiting competent members to the Comittee. The appointment of each member should include an understanding of the limits each would set on his time availability."

Response

We agree, with coment.

The average service for members is about 120 days per year.

Those who can spend only significantly less time than this should not be ruled out for consideration, but we believe that a minimum of about 80 days per year is necessary to make a sustained contribution to the work of the ACRS. We intend to use these guidelines, flexibly and informally, in evaluation of potential members.

Recomendation C.10. "The specific composition of the Comittee as measured by the skills represented may be reasonable for.. the present, but we believe that the composition should be changed as the direction taken by the Committee changes. For instance, the growing im-portance of waste management implies that a member should have a

, background in chemical process engineering. Additional fonner senior utility management with experience in nuclear plant op-erational management would also be desirable, as well as members with NSS'; systems design experience."

i 101

i

, 1

Response

We agree, with comment.

We agree with the concept that the mix of technical disciplines represented in experience of the members should change as the )

mission of the NRC and the ACRS changes. We do not necessarily <

agree with all of the example suggestions made by the Panel. We believe it is important to recognize that the mix can be changed only slowly and with consideration of other factors, some of which we believe are more important [see C.8. and C.9.].

Recommendation C.11. "The list of ACRS consultants is very long, but most are not used very much. We do not find the use of consultants to be excessive. The Comittee should be careful to ensure that con-sultants do not appear .to be speaking for the Comittee. Con-sultants should only advise the ACRS."

Response

We agree, with coment.

We note that our use of consultants over the past few years is

decreasing and we expect the trend to continue.

Recomendation a

C.12. "The size and composition of the ACRS Staff should be reexamined once the ACRS mission is changed as we have recomended, and assuming this change takes place."

Response 1 l

We agree, with coment.

We have accepted the recomendation, largely under the pressure of budget restrictions. The total of personnel assigned to ACRS will be only 42 in FY 87 compared to 54 two years ago. We are accomodating the reduction largely by reducing service staff rather than by reducing the number of members, consultants, or

! technical staff. Some reduction in technical programs will also be necessary, and we are assigning priorities to all current l

programs, considering importance to nuclear safety, and, in some cases, other Comission requirements. We have concern that too great a reduction in staff support will cause the Comittee to be less effective and make membership professionally less attractive for present or potential members. Therefore, we do not intend to make further reductions in staff without comensurate reduction in programs.

102

Recommendation C.13. "A channel should be established for ACRS to submit users' requests for research to the Office of Nuclear Regulatory Re-search, just as is done by the Regulatory Staff."

Response

! We disagree with this recommendation. We can adequately influ-ence research programs under the present system and oppose ac-quiring the responsibility and bureaucratic load involved in becoming a formal user office.

Recommendation C.14. "ACRS should establish rules to cover the circumstances in which d.iscussions between individual ACRS members and individual Commissioners can take place, and the protocol for these dis-

] cussions."

Response

j See response to B.S.

I 1 Recommendation C.15. "The current conflict of interest and FACA requirements are not unduly restrictive."

Response

We agree, with an exception.

We agree, so far as FACA requirements, but disagree with regard 1 to conflict-of-interest requirements. We believe the interpreta-tions of regulations in this latter area for application of the activities of ACRS members may be unnecessarily restrictive. We are concerned that activities of members can be arbitrarily and unfairly restricted, and believe the matter warrants reevalua-tion.

I 103 i

/ o UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

1. E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS k #

WASHINGTON, D. C. 20555 4 ~... /

July 15, 1986 Honorable Lando W. Zech, Jr.

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

Dear Mr. Zech:

SU3 JECT: ACRS COMMENTS ON THE DRAFT COMMISSION POLICY STATEMENT ON TECHNICAL SPECIFICATIONS During its 315th meeting, July 10-12, 1986, the Advisory Committee on Reactor Safeguards reviewed the Draft Commission Policy Statement on Technical Speci-fications transmitted to us by a memorandum from William T. Russell, NRR, dated June 30, 1986. This matter was also discussed during a meeting of the ACRS Subcommittee on Plant Operating Procedures on July 1, 1986.

In our view, the intent of the proposed statement is appropriate. We agree that the technical specification requirements and NRC practices have resulted in a large number of detailed technical specifications that, in some cases, place an unnecessary burden on licensees and plant operators. Furthermore, we welcome the effort made to clarify the scope and purpose of technical specifications and commend the NRC Staff on its work with the Atomic Industrial Forum in developing a set of criteria for screening items for inclusion in revised technical specifications.

We recommend that the Commission issue the Draft Commission Policy Statement for public comment. We would like to meet with the NRC Staff again on this issue after comments by the public have been received and taken into con-l sideration.

Sincerely, I

David A. Ward Chairman

Reference:

Federal Register Notice for " Draft Commission Policy Statement on Technical Specification Improvement," Draft 2, transmitted by memorandum from William T. Russell, NRR, to Carlyle Michelson, ACRS, dated June 30, 1986.

105

REVISED 8/15/86:

d p urg'o UNITED STATES Additional Coments, P.4 8 o NUCLEAR REGULATORY COMMISSION W E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

1. WASHINGTON, D. C. 20555 s,o m, / August 12, 1986 Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555 1

SUBJECT:

ACRS COMMENTS ON PROPOSED NRC STANDARDIZATION POLICY STATEMENT During its 316th meeting, August 7-9, 1986, the Advisory Committee on Reactor Safeguards reviewed the Comission draft of April 10, 1986 and the NRC Staff response of May 14, 1986 on the Proposed Standardization Policy Statement as requested by Chairman Palladino's memorandum of June 18, 1986. In this review, we had the benefit of briefings by the NRC Staff during the 315th ACRS meeting, July 10-12, 1986, as well as during the meetings of our Subcommittee on Improved Light-Water Reactor Designs on March 12, 1986 and August 5, 1986.

We agree that standardization of nuclear power plant designs is desir-able and may lead to enhanced safety if properly implemented in accor-dance with other Comission policies, particularly those on safety goals and severe accidents. For this reason, we believe that the plans and directions for implementation that will be provided in a forthcoming Staff report (NUREG) are crucial. We expect to review this report when it becomes available.

As the result of our review, we offer the following detailed coments and recomendations:

1. We recomend that the title of this policy statement be changed to

" Policy Statement on Certification for Nuclear Power Plant Standard Designs." We believe that the policy statement should focus on standardization of the design of nuclear power plants. We do not recomend including in the policy statement a coment on standard-ization of procurement, constructien, installation and quality assurance practices, training and emergency operating procedures, or maintenance procedures. To require standardization of these items would be overly prescriptive, although certain elements of these practices and procedures will become standard as a beneficial result of the standardization of design.

2. It is our opinion that the first sentence of the Commission's draft I policy statement dated April 10, 1986 best represents the Com- l mission's policy on certification of nuclear power plant standard I designs.

l l

107

Honorable Lando W. Zech, Jr. August 12, 1986

3. We do not recomend including a coment in the policy statement "that standardized nuclear power plants should be used to satisfy the ultimate licensing goal of certified designs constructed on preapproved sites." This is an overly restrictive statement of the purpose for standardized nuclear power plants.

4. We believe the policy statement should make clear that this state-ment supersedes the Comission's previous policy on standardization issued in 1978. This is necessary because the 1978 policy contains 4

obsolete provisions and does not reflect present Comission regu-lations and policies.

5. Considering the above coments, we submit for your consideration the following revision of the heading and the first two paragraphs of a proposed policy statement:

POLICY STATEMENT ON CERTIFICATION FOR NUCLEAR POWER PLANT STANDARD DESIGNS The Nuclear Regulatory Comission believes that standardiza-tion of nuclear power plant designs is a very important initiative that has the potential for significantly enhancing the safety, reliability, and availability of nuclear plants.

The focus of this policy is the reference system design certification. The goal of standardization should be an essentially complete plant design with respect to both scope and level of detail, which then can be referenced in indi-vidual license applications.

This policy statement supersedes the Comission's previous

" Statement on Standardization of Nuclear Power Plants,"

, published August 31, 1978. Details of the issues and topics that are important to the execution of this policy and other 1

short-tem licensing transition options are discussed in NUREG-XXXX, including the definition of scope and detail of an essentially complete plant design required for certifica-tion. Applications not referencing -a certified design will be processed in accordance with existing Comission regu-lations and policies, as discussed in NUREG-XXXX.

6. We recomend including in the policy statement a reference to Comission policies on safety goals, severe accidents, and advanced riactors, as well as reference to other Comission policies pertinent to future nuclear plant designs and the manner in which the requirements of these policies in future designs should be defined in the accompanying NUREG.

7. Fonner Chairman Palladino's coments regarding the need for empiri-cal information on safety features that differ from those in 108

) __

Honorable Lando W. Zech, Jr. August 12, 1986 existing plants and for prototypical tests of " entirely new de-signs" raise questions of considerable interest and importance. We are not prepared to comment on these matters at this time, but expect to consider them further and to discuss them with the' NRC Staff during our consideration of the anticipated Staff NUREG.

8. It is our understanding that design certification rulemaking options will be discussed ir the proposed companion NUREG. It is

- not clear whether that NUREG will be published for coment; there-fore, it is not clear if or at what time the public will have an opportunity to provide coment on these options. 'We believe that the Commission would benefit from and should seek public coment on design certification rulemaking options.

If informed coment is to be obtained, we think that the criteria and thresholds for standing and interest for participation in the legislative or adjudicatory rulemaking hearings should be made clear.

9. The proposed outline of the NUREG appears satisfactory. However, it is important that the definition of " essentially complete" design be thoroughly and clearly identified as to the complete scope and level of detail of information required for design certifications. It is also important that the scope and level of detail be equally identified for each of the other options.

With the above coments and recomendations taken into account, we believe that the policy statement should be issued for public coment.

Additional coments by ACRS members David Okrent and Glenn A. Reed are presented below.

Sincerely, David A. Ward Chairman Additional Comrents by ACRS Member David Okrent I wish to indicate first that I an a strong supporter of standardiza-tion. In fact, I would take such steps as are legal to limit severely the nusber of certified standard reference designs to be approved by the NRC.

Second, I wish to support fomer NRC Chairman Palladino in his position that standardization should ideally encompass essentially complete design of the entire plant and that empirical information or proto-typical testing of new features is important for certified reference plants.

109

Honorable Lando W. Zech, Jr. August 12, 1986 Third, I believe that future U.S. plants should be considerably improved in safety over current U.S. plants and that this should not be left to the whims of the designer or the vagaries of PRA. I believe that the Comissioners should explicitly state that they will seek a higher level of safety and that specific safety features and performance goals are to l be included in the design of certified standard reference plants. These safety features and goals would best be specified prior to adoption of a new standardization policy statement.

Additional Comments by ACRS Member Glenn A. Reed In my opinion, the Policy Statement on Nuclear Power Plant Standardiza-tion should include statements beyond the standardization of an "essen-tially complete plant" in order that the several different plants that are likely to result become more standardized in key safety features and systems. There are elements of imaturity and differences in safety systems of the PWRs of the different vendors which should be trending toward sameness. I consider the policy, as now written, will not encourage, at a satisfactory pace, the standardization of these systems of the " essentially complete plant."

References:

1. Memorandum dated April 10, 1986 from Samuel J. Chilk, Secretary, to Victor Stello, Jr. , Executive Director for Operations,

Subject:

Standardization Policy Statement

2. Memorandum dated May 14, 1986 from Victor Stello, Jr. , Executive Director for Operations, to Comissioners,

Subject:

Standardiza-tion Policy Statement Revised Paae

~

110

/ 'o g UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o wAssincrow. o.c. zossa

• • • • • October 15, 1986 Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS COMMENTS ON DRAFT NUREG-1225, " IMPLEMENTATION OF NRC POLICY ON NUCLEAR POWER PLANT STANDARDIZATION" During its 318th meeting, October 9-11, 1986, the Advisory Comittee on Reactor Safeguards discussed the referenced draft NUREG-1225, "Implemen-tation of NRC Policy on Nuclear Power Plant Standardization." A sub-comittee meeting on this matter was held in Washington, D. C. on October 8, 1986.

We are in general agreement with the draft NUREG-1225, but we have the following observations:

1. It is not clear that the proposed NUREG-1225 will be published for public comment. We believe that the Comission would benefit from and should seek public coment on the design certification rule-making options. Also, if informed coment is to be obtained during the rulemaking process, we think that the criteria and threshold for standing and interest for participation in the legislative or adjudicatory rulemaking hearings should be made clear. We under-stand that the provisions for participation will be defined in the notice of rulemaking in the Federal Register for the specific rulemaking proceedings; this should be so stated in draft NUREG-1225.

2. We do not consider that the scope and level of detail of infoma-tion required for design certifications are adequately defined in draf t NUREG-1225. It should be made clear that, in addition to providing a level of design detail equivalent to that required by 10 CFR 50.34(b) for a final safety analysis report, an applicant for a final design approval (FDA) should be prepared to supply such other information as is customarily required by the NRC Staff to perform a final safety analysis report review.

Since an FDA for a final design must be issued before the design I can be certified, the certification process ideally should require i little additional design information if that supplied with the FDA is adequate. However, the scope of design presently described in Section 3.1.3, " Design Certification Concept," of draft NUREG-1225 is not adequate and needs to be expanded and better defined.

111

Honorable Lando W. Zech, Jr. October 15, 1986 i

We believe that the expansion and clarification of infomation requirements for an " essentially complete design" should have input Srom the principal cognizant NRC Staff reviewers and various

industry organizations experienced in such matters.

3. It should be made clear that portions of a desion which has re-ceived design certification by the NRC are not thereby certified for other applications.

The ACRS would like to be kept informed regarding this matter.

Sincerely.

00R~Q David A. Ward Chaiman

Reference:

Draf t NUREG-1225, " Implementation of NRC Policy on Nuclear Power Plant Standardization," undated, Handout during 318th ACRS meeting, October 9-11, 1986 l

I 112

l l p**8%q'o,,

~

UNITED STATES i P n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAssincron. o. c. zossa g

October 15, 1986 o....

Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS SUGGESTIONS FOR AN NRC LONG RANGE PLAN As a follow-up to our letter of August 13, 1985, we have devoted considerable effort to the development of our thoughts and recommenda-tions on a Long Range Plan for the Nuclear Regulatory Ccr.ission.

Included in this effort have been discussions with a dczen highly respected people knowledgeable in the nuclear regulatory e.rea. This effort has resulted in the attached material which constitutes a series of suggestions for development of such a plan, and we forward it for your consideration and use. We recognize the NRC Staff effort directed toward development of a long range strategy for the Agency, and we hope that our report will be useful in that effort.

Sincerely,.

d

. N .

David A. Ward Chairman

Attachment:

ACRS Suggestions for an NRC Long Range Plan i

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i October 15, 1986 l

ACRS SUGGESTIONS FOR AN NRC LONG RANGE PLAN

1. INTRODUCTION

Background:

, The ACRS concluded during November 1984 that the NRC should have a Long Range Plan and that it could assist the Comission in developing such a Plan. The Committee subsequently appointed a subcomittee which held a series of nine meetings, some to deliberate and some to seek infomation and advice from a ,

group of senior, respected people who are knowledgeable in the nuclear regulatory area. These people were:

1 o John F. Ahearne, Vice-President of Resources for the Future

and former Chairman of the U. S. Nuclear Regulatory Comission, i o Peter Bradford, Chairman of the Public Utility Comission of Maine and former Comissioner of the U. S. Nuclear Regulatory Comission, o Floyd Culler, President, Electric Power Research Institute, o Carl Giesler, Vice-President, Wisconsin Public Service Corporation, I o James Knight, Director, Licensing and Regulatory Division, U. S.

Department of Energy, (Acting Director, Division of Engineering, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission,

! wheninterviewed),

o Roger J. Mattson, Vice-President, International Energy Associates, Ltd. and fomer Director of the Division of Systems Integration, Nuclear Reactor Regulation, U.S. Nuclear Regulatory Comission, j o William McDowell, NRC Audit Group, General Accounting Office, j o Warren Owen, Executive Vice-President, Duke Power Company, o Victor Stello, Jr., Executive Director for Operations, (Deputy j Executive Director for Regional Operations and Generic Require-I ments, U. S. Nuclear Regulatory Comission when interviewed),

1 o John Taylor, Vice-President, Electric Power Research Institute,

o James Tribble President, Yankee Atomic Electric Company, i

j o John West, Vice-President (retired), Combustion Engineering Corpo-ration.

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2 Written transcripts were made of most of the discussions with these people, and we comend them to your planning staff.

The subcommittee's efforts have produced two primary results, the first being a letter to Chairman Palladino dated August 13, 1985 in which the ACRS formally stated its belief that the NRC should have a long Range Plan. The second result is a group of suggestions for the development of an NRC Plan.

These suggestions were considered by the Committee during its 316th: 317th, and concluded during its 318th meeting, October 9-11, 1986. We wish to call them to the Commission's attention for use in the planning effort. We believe they will assist in providing a foundation on which the Commission can develop a useful and needed Plan.

Scope:

Although any NRC Plan would likely specify actions for only about a five year period, it would need in some parts to be based on assumptions of events or circumstances ten or twenty years into the future. In addition, some goals and objectives may require considerably longer to accomplish than five years even though it may be impossible to define specific actions to be taken beyond tirat period of time. We have given recognition to these facts.

This study primarily addresses long-range concerns which may be receiving little of the Agency's attention rather than today's problems which normally receive heavy emphasis.

Since this is an initial effort, we have restricted our advice primarily to nuclear power plant and high level waste (HLW) issues and have paid little or no attention to other areas such as nonpower reactors and radioactive mate-rials 11 censing.

Outline:

In compiling our suggestions, we have chosen to utilize the following out-line:

1. Introduction

2. Mission Statement: The mission of the Agency as we perceive it is defined.

3. External Assessment: We have attempted to list most of the ma,ior factors which we believe will describe the likely regulatory environment in which the Agency must operate for the next few years.

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4. Internal Assessment: We have tried to enumerate the major strengths and weaknesses of the Agency as we and many other people view them,

5. Possible Changes by the NRC: We have given many examples of possible, important changes which could be considered in an effort to enhance the Agency's performance and to prepare it to meet future problems and challenges.

6. Goals and Objectives: We stop short of designating goals and objectives, believing that this step is the responsibility of the Commissioners themselves. We view the material presented as a foundation upon which the Goals may be based.

2. MISSION STATEMENT This section deals with questions such as: What does the NRC do? What are its responsibilities? What did the Congress intend for it to do? In our judgment, what should it do?

Section 1 of the Atomic Energy Act states: "It is - declared to be the policy of the U. S. that (a) the development, use, ai d control of atomic energy shall be directed so as to make the maximum contribuuon to the general welfare, subject at all times to the paramount objective of making the maximum contribution to the common defense and security; and (b) the development, use, and control of atomic energy shall be directed so as to promote world peace, improve the general welfare, increase the standard of living, and strengthen free competition in private enterprise."

We believe that the major responsibilities of the NRC with regard to nuclear power plant regulations and waste management t.re:

2.1. First and foremost, to provide a regulatory environment which will limit to an acceptable level the probability and consequences of any nuclear power plant accident which would have a major impact on the public health and safety (including in-plant personnel).

2.2. To assure that routine operation of nuclear power plants does not result in an undue impact on the public health and safety (includingin-plantpersonnel).

2.3. To meet the above responsibilities while imposing the minimum regulatory and economic obstacles on the development of the uses of nuclear energy for the benefit of the nation.

2.4. To determine the levels of safety which are appropriate to the public interest.

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4 2.5. To assess the degree to which its objectives are being met.

I 2.6. To enunciate clearly to the public the objectives and results of i its efforts. (To regulate in a fashion which will give the public

' confidence in the regulatory process.) ,

2.7. To accomplish its mission efficiently.

Major activities of the NRC are or should be:

2.8 To issue safety approvals and licenses, write and set standards and criteria, lead in the development of national nuclear safety goals, t

and guide the development of nuclear technology in a safe manner.

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2.9 To advise other federal and state agencies and groups on these and related matters.

2.10 To integrate its nuclear safety efforts with those of other govern-3 mental, private, and international groups.

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NRC actions and/or approvals have considerable influence on:

4 2.11 The regulated industry.

2.12 Competition among vendors, sub-tier vendors, countries, and so on.

4 2.13 Actions by intervenor groups.

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2.14 State legislatures and regulatory agencies.

2.15 Actions by government agencies, safety organizations, and vendors i in other nations.

l 2.16 The public view of nuclear power plant hazards and public health l and safety.

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The NRC is distinguished by characteristics such as the following:

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2.17 The NRC was granted federal preemption in radiation safety matters I by the Congress, and the courts have upheld this preemption.

2.18 The NRC has lengthy nuclear regulatory experience and a strong institutional memory, and it has contributed to an excellent record

in terms of protecting the public health and safety.

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5 2.19 The NRC has the ability to cause major perturbations, for better or worse, in the industry it regulates and in the cost of electric power to the consumer. '

3. EXTERNAL ASSESSMENT The Long Range Plan must relate to and be consistent with the societal, I economic, and technical environment of the time period for which the Plan is developed. This environment must be defined as well as possible before the Plan is developed, and the Plan should be changed, as needed, if the environment changes. We believe that the factors listed below are some that should be considered in attempting to describe the environment:

Current Situation 3.1. Construction of approximately 120 nuclear power plants will have been completed in the next five years or so, and the utilities will endeavor to operate many of these plants for lifetimes of 50 to 100 years.

3.2 It is extremely unlikely that any new energy sources for the large-scale production of electricity will be demonstrated to be economically competitive during the next decade.

3.3 Efforts to conserve energy will become less effective in diminish-ing the demand for more growth in electricity production capability as the easier steps are finished and the more difficult ones are undertaken.

3.4 The cost of electricity from coal will continue to increase as the impact of fossil plant effluents on the environment and climate, the impact of S07 et al on public health, coal mine safety, and the disposal of fossT1 wastes receive more public recognition. For these same reasons, the cost of electricity from oil and gas will also increase, although not as markedly.

3.5 The use of oil to generate electricity will continue and perhaps increase modestly over the next five to ten years due to recently lowered petroleum prices. However, there is also considerable possibility that prices will rise significantly during this period and even the possibility that a Mideast crisis will occur which will deny oil to the West. The need for nuclear power would increase in either case, quite drastically in the latter.

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Factors Delaying New Nuclear Power Plant Orders 3.6 There may be a modest shift back to the use of natural gas to generate base-load electricity, and such usage may continue over the next decade.

3.7 The cogeneration of electricity will increase nationally and retard the need to order new, large coal or uranium plants by as much as l

j five years in some parts of the country.

3.8 The date for ordering new power plants will be delayed somewhat as relatively modest amounts of power are purchased from Canada or

' Mexico. This date will also be delayed if Congress passes legis-lation to permit utilities to use existing transmission lines more easily to sell power to other utilities in other regions of the country or if it passes legislation to establish a national grid.

3.9 In the present political and economic climate, most utility com-

, panies will delay the construction of large-scale plants as long as possible. This will result in the use of short-tenn-economic but long-term-uneconomic solutions to meeting power needs (such as through the use of low-capital-cost, high-fuel-cost, high-cost-per-kw-hr gas turbines) which will frequently not be in the nation's best long-term interests.

i 3.10 The unlikely prospect for new nuclear power plant orders during the next few years stems largely from matters such as:

a, uncertainty over probable State regulatory actions in

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ensuring adequate rates of return; e.g., inability to charge construction-work-in-progress costs,

b. uncertainty over the impact of a new nuclear commitment on the financial rating of an electric utility,

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c. uncertair y over the economic effects of possible NRC backfit requirements,

d. uncertainty over the lack of a national consensus on the need for nuclear power,

e. uncertainty over the time required to license, construct, i

and place a nuclear power plant in operation,

f. uncertainty over the status of the Price Anderson pro-vision, and 119

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'; g. uncertainty in estimating the growth in electricity '

usage.

Factors Significant to New Construction Activities I 3.11 It is unlikely that any utility will place an order for a new nuclear power plant within the next few 3 ears. It is quite possible, however, that a utility or a consortium may apply to reinitiate construction on a partially-constructed-but-deferred plant within the next five years.

3.12 Regional nuclear-power-generation companies which will generate electricity and sell it to utilities for distribution may come into being. State public utility commissions may encourage such com-panies in the same context as cogenerators are encouraged.

3.13 It is believed that utilities will lean very heavily toward pur-chasing advanced LWRs which have evolved from present-generation LWRs when they purchase new plants. Emphasis may be placed on small (i.e., 500-600 MWe) as well as on large plants. They will give serious consideration to duplication, replication, and stan-dardization of plants.

3.14 The cost of electricity from uranium in t/ kwhr would be signifi-cantly less from a modern nuclear plant constructed "today" than from a modern coal plant if the nuclear plant construction time

, could be reduced to a five-or-six-year period. Such a construction time can be achieved by competent utility management if the nation-al and state regulatory processes can be improved. In fact, a i

representative from one utility company stated it can build nuclear plants today from which electricity costs would be 20% lower than from coal plants.

3.15 It is believed that new nuclear power plants will be ordered in as short a period as threr. years or as long as 15 years from now. The actual date will depend largely upon economic matters such as:

the cost of electricity from coal 4

o o the cost of oil o the cost of money However, it will also depend on social factors such as political pressure on utility organizations to shift from ensuring conserva-tion to ensuring the c.vailability of low cost electricity when needed. Earlier dates will become more likely as plant con-struction times decrease, interest rates decrease, regulatory 120

8 uncertainties decrease, the need for electricity increases, and so on.

3.16 It is expected that several foreign nuclear power plant vendors (e.g., from the Federal Republic of Germany, France and Japan) will be ready to make serious, full-fledged sales efforts to U.S.

utilities within five years. It is expected that advanced LWRs will have been built and be operating in foreign countries within a decade. These plants will have been licensed under regulatory programs based generally on the U.S. model, will have been constructed in five-year-or-so periods, and some will be producing electricity quite economically. U.S.Inutilities will be strongly addition, there is attracted to purchase these plarli.

considerable likelihood that U. 5. manufacturers will have lost much of their nuclear power plant design and construction capability within the next five years; they should have retained capability to service existing plants, but they may well have lost the capability to produce and supply major components for new plants (e.g., pressure vessels) without strong support from foreign affiliates.

Factors Influencing Nuclear Power Plant Safety 3.17 Some nuclear power plant utilities may go bankrupt during the next decade for many reasons including the possible advent of deregulation of the utility industry and the associated development of competition among utilities. The potential impact of such developments on the implementation and utilization of safety procedures will have to be considered.

3.18 State public utility commissions will, in some cases, adversely affect nuclear power plant safety as they impose economic incentive programs, limit utility research, and so on.

General Matters 3.19 With 120 nuclear power plants soon to be in operation, one can anticipate a continued occurrence of abnonnal events. Some of these may cause potential adverse health effects or even deaths among the plant labor force or even among neighboring populations.

Any such event will undoubtedly cause considerable alarm among members of the public (although perhaps not as much as occurred at THI), and the NRC credibility will suffer badly unless positive steps have been taken to inform the public that such events can occur. This loss of credibility could frustrate the objectives of the Atomic Energy Act.

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9 3.20 The NRC must provide assurance that aging nuclear power plants can continue to operate safely.

3.21 Little decomissioning of nuclear power plants will occur in the next decade or two except for early, small plants. There may be demands for more assurance that plants can be decomissioned economically, but the responsibility for showing this will fall largely on the utilities or DOE and not on NRC.

3.22 The U.S., under DOE leadership, will continue development work on advanced liquid-metal and gas-cooled reactors. It is possible that the U.S. will initiate design of an advanced plant within five years and that the NRC will be called upon to license it.

3.23 Legislators and the public will continue for another decade to view HLW disposal as a serious problem. The national repository disposal program will fall behind and target dates will not be met, increased emphasis will be placed on temporary, above-ground storage, and the NRC will need to continue to give high priority to HLW activities as at present. A similar problem exists relative to the disposal of low level waste.

3.24 Serious efforts to initiate fuel reprocessing will not be made in the next decade because of marginal economics as well as safeguards concerns.

4. INTERNAL ASSESSMENT This section is directed toward defining the strengths and weaknesses of the NRC. Among the strengths of the Agency are:

4.1 It has a fine track record in protecting the national public health and safety. Further, it has had a strong influence on the regulatory programs in most of the Western countries, and the records for protecting public health and safety have been outstanding there, also. It deserves its fair share of the credit for this additional accomplishment.

4.2 It has extensive knowledge of the design of safety-related systems. I The NRC, as a group, has a strong knowledge of how designs have  !

changed with time and the accompanying reasons. This knowledge centers on the safety aspects of the designs rather than on the economic aspects.

4.3 It has a good understandirg of risk assessment and its related applications in the nuclear power area. There probably isn't any 122 '

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) better group in the world than the NRC Staff and its contractors in *

,i the field of risk assessment.

! 4.4 It has a significant institutional memory and a knowledge of good ,

i engineering practice. The Agency has key people and groups which have been involved in regulatory-related activities for two decades or more. Individual members have obtained a strong knowledge of engineering practice both before and after joining NRC. Participa-tion by members in the codification of national standards (e.g.,

ASME piping code) has been a significant factor here.

4.5 The NRC has recently completed and published a policy statement on safety goals for the operation of nuclear power plants. The Agency l 2

l has also formulated a Severe Accident Policy Statement. The

development and formal issuance of these two documents represents a j signal accomplishment on the part of the NRC and makes it the only Federal agency that has directly addressed and publicly enunciated policies on such controversial issues.

j 4.6 It has a competent technical and administrative staff.

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Among the weaknesses of the Agency are:

! 4.7 It is lacking in the area of long range planning. This was demon-strated by the fact that the Agency was totally unprepared for the i

Three Mile Island accident, although considerable progress has been i

made since then.

I I 4.8 There are fundamental questions in terms of the way the Commission is set up and operates:

o There is a lack of direction within the Agency. The planning process, no matter how well done, will not be executed well because the Commission has trouble, first, in articulating its  ;

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priorities and second, in getting them implemented.

! o Information does not flow well within the agency. Many components of the Agency really do not comunicate well with one another, largely because communication channels have

' become unduly formalized. People in one part of the Agency tend not to talk to people in another part. There is some-j times a tendency for people not to work together to solve I common problems. The Commissioners seem frustrated about their lack of ability to share ideas among themselves, to

! ventilate issues, and to talk on a professional level with the i

Staff because of the applicable statutes.

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11 4.9 It has a tendency to regulate in an economically wasteful fashion.

NRC must accept its share of the blame that backfitting has been necessary and very expensive, that QA/QC requirements lead to very high costs, that " nuclear power has priced itself out of the market" in the words of one utility representative, and so on. In some cases it has established undol sometimes unnecessary regulationse.g., (y conservative, 10 CFR Part expensive, 21-Reporting ofDefectsandNoncompliance). The Staff builds in conservatism

discipline by discipline and parameter by parameter, rather than being as realistic as the technology will support and then adding an appropriate factor of safety at the end. It seemingly does not act on the results of safety research (if in the direction of 1

' relaxation), and requirements once established are seldom relaxed.

The Staff seems frequently not to arrive at practical, cost-effective ways to solve problems.

' 4.10 There is an adversarial relationship between the Agency and the regulated that sometimes detracts significantly from the ability of the Agency to operate in an effective and efficient manner.

4.11 It lacks the ability to regulate in a fashion which builds confi-dence and trust in the regulator. It has not exercised adequate discretion in avoiding relatively needless modification and back-fits, although the Requirements (CRGR) has performance of the Comittee been very encouraging. to Review Generic (There is a need to bring more operating experience to the Staff and to improve the

Agency's understanding of the operational nuances of the plants that it regulates.) We are told by industrial representatives that industry will not build new plants until it has confidence that a

stable, predictable regulatory situation exists.

4.12 It has developed a regulatory system that is so comprehensive, and frequently so prescriptive, that both the NRC and many of the l operating utilities have come to believe, or act as though they believe, that compliance with the regulations is itself sufficient to assure safety. The assumption that regulations and safety are synonymous may be dangerous and should be reexamined.

4.13 With the new Safety Goals, the Comission has addressed the ques-tion, "How safe is safe enough?" However, it has not decided what it intends to do with the Goals -- how it intends to implement them. It has not resolved the fundamental questions: "Do we want to make operating plants safer than they are now? If so, how much?

How do we decide?" For example, how ought the unresolved safety issues and the elements of the Severe Accident Policy Statement be handled?

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! 12 4.14 It has not established a convincing argument for how much and how long it should continue doing safety research.

4.15 The Agency has an inadequate public information program. The public is informed of public health and safety hazards related to nuclear power and of the NRC's regulatory activities only as information comes out through the media, which are themselves poorly educated and frequently biased. The NRC has no program of going out and truly informing the media especially and the public in general. The quality of reporting of the next major nuclear accident will depend upon the quality of the relations between the NRC and the media, which will in turn depend upon the credibility that is established during the preaccident period. This can have a profound effect on the degree of panic engendered by the media, with inevitable consequences for the public health and safety.

4.16 The Agency has no assured supply of new blood. It has no plan to assure that technologists will be available as needed in the future and that relevant academic programs will survive the quiescent period of the next few years.

4.17 The Agency has no Congressional or public constituency. No one gives it encouragement, and that probably affects what it does.

Outsiders may view the Commission and Staff as vacillating as the Agency is criticized from all sides and as it attempts to satisfy all sides.

4.18 Much of the public criticism of the NRC is ill founded and lacks credibility, but some is constructive. The Agency tends to respond i

to both kinds by becoming defensive in its reponse.

5. POSSIBLE CHANGES BY THE NRC This section deals with development and decisions on alternatives. What could the NRC do differently? What could it change? What is it possible for  !

the NRC to do in each of the areas above where weaknesses, strengths, oppor-tunities, threats, current capabilities, and so on have been pinpointed?

(Note: We are interested primarily in what the possibilities are with regard to correcting potential or actual weaknesses, and the following is prepared in that context.)

The NRC 5.1 Could initiate a long range planning activity which would incorpo-

.l rate not only forecasting and trying to look ahead but also policy setting and positioning and preparing things for the future -- not just forecasting things beyond its control and attempting to

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13 accomodate them but actually trying to shape the future and developing a strategy for the future. This might be termed strategic planning rather than long-range planning. It would look farther and more broadly into the future than the Policy and Planning Guides (PPG), and the latter could likely be a subset of a long Range Plan.

5.2 To study its internal operation, could engage an outside firm of management consultants to review the way the NRC conducts its business and to recommend changes as appropriate.

5.3 To regulate in a more cost-effective fashion, could do such things as:

a. Set up an imsartial, learned group to study whether regulation in the U.S. aas been unduly costly. Such a group could examine allegations such as those in a recent Booz, Allen, and Hamfiton, Inc. study which states that regulation caused over 70% of the nuclear plant cost escalation in the U.S. between 1973 and 1983, that nuclear building costs grew 22% per year vs.12% per year for coal plants during the period, and that cost growth in both France and Japan was at a rate much closer to the rate of U.S. coal construction growth than the rate of U. S. nuclear growth. Such a group could compare regulatory cost effectiveness in the three countries taking into account the strong similarity between safety standards and practices,

b. Undertake a study to determine if the NRC follows up the results of research work to tiohten or relax standards and requirements as appropriate. If not, it could institute and enforce a policy to correct this situation,

c. Make a specific attempt to establish a consensus among the NRC, the Congress, state public service commissions, industry and other special interest groups, and the general public as to when the expenditure of public resources (money) to achieve higher levels of public health and safety is no longer appro-priate. Such a consens*4s could be based on the Safety Goal Policy Statement.

5.4 To minimize the adversarial aspects of regulation, could develop methods to place greater reliance on industry self-policing initia-tives or to try harder to instill in utilities the desire to do a better job and to seek excellence. The activities of the I1stitute of Nuclear Power Operations are supportive in this area.

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14 5.5 To endeavor to remove the large uncertainties (actual or perceived) from the licensing process, to bring greater stability to the regulatory process, and to establish a positive climate for the i renewal of growth in the nuclear power industry, could do such things as:

a. Establish a clear distinction between the safety requirements for new plants vs. existing plants. (Althoughautomobilesare recalled to repair defective brakes, they are not recalled to replaceoperablebrakeswithimprovedones.)

b. Consider an alternative approach to present regulatory trends wherein the NRC is moving in the direction of giving greater attention to the balance of plant. One alternative to that approach could be to focus regulatory activities on a defined

" safety envelope" of systems and equipment which would address only the essential core-safety and public-safety requirements.

For example, for LWRs, it might be possible to limit the safety envelope to a highly reliable:

- Core subcriticality mechanism.

- ECCS and decay heat removal system,  ;

- Reactor vessel and primary piping with inspectability throughout service life, and

- Containment, coupled with a performance-based limitation on the frequency of safety challenges from the balance of plant.

The intent would be to specify only the systems and components absolutely required for safety, and only these would be rigorously regulated. Other systems and components throughout the plant would be under the utility self-policing initiative and Institute of Nuclear Power Operations good practice standards.

An integral part of this approach might be the willingness of the NRC to answer the question, "If we don't think it is safe, what will it take to make it safe enough?" If an envelope as proposed does not meet the requirements in NRC's judgment, the

, Agency should be willing to state what would meet them. The NRC would not design components of the plant; rather it would say, "There are probably many approaches that would satisfy requirement X. One such approach is such-and-such providing all the detailed considerations are handled adequately."

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c. Influence plant design to minimize the impact of off-site hazards and to reduce the size of emergency planning zones.

d. Continue pursuit of the source term work with the objective to more properly define emergency planning zones.

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e. Continue the recently initiated efforts to interact with the designer on licensibility matters in the early design stages of advanced plants.

f. Develop improved General Design Criteria based on experience of the past twenty years, including insights gained fron, improved methods for risk and consequence analysis,

g. Orient some of the NRC LWR research programs toward " creative research" or " exploratory research" -- looking for new ways to bring about better safety. Perhaps the Department of Energy should be the group to do this, but there is little evidence that it does. ,

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h. Position itself, the industry, and the nation so that, when another major accident such as TMI occurs, the nation will not lose the benefits of nuclear energy. This could include activities such as:

o Development of an effective program to educate the media on nuclear power plant safety and on the concepts of risk and probability, o Steps to explain to the public the kinds of efforts that are made to ensure public health and safety, to compare for the public the safety aspects of electricity from uranium to those for electricity from other realistic energy sources, and to explain to the public the benefits and costs of nuclear power, o Efforts to make the public realize that there could be another accident involving severe core damage and that it should not mean the end of nuclear power.

5.6 To be prepared to carry out changing missions in the future, could do such things as:

a. Analyze whether it has the right mix of technical people to accomplish its probable future missions, and how it can keep those people active in any period of minimum activity. Can technical people expert in one area today be shifted to other 128

16 work temorrow and still be able to shift back five or ten years later? Can design analysts be effectively shifted to inspection-oriented activities? Should the NRC reduce to a minimum staff level soon, with the expectation of expanding later? Will the appropriate people be available on the market later? Does the NRC need a more extensive and more formal internal or external continuing-education program? )

How will the NRC retain its share of the more experienced people and not be left with a stable of less competent people for the time when there is a resurgence in the licensing of new designs? Can it be done? How can the NRC add more "workplace experience" to its Staff?

b. Consider how it will regulate on a (likely) declining budget and how it will conduct the necessary research on a (likely) declining budget.

5.7 Could give consideration to a wide variety of important matters which do not fit in the categories above, such as:

a. The ramifications of the Agency changing from an organization heavily involved in licensing to one which is primarily an inspection group. Several hundred people may be switched from the design / construction area to the operational area, and careful control may be required to prevent the establishment of large numbers of unneeded positions and a significant loss in agency efficiency,

b. Whether the NRC should license the utility as it does now or whether it should license the plant (as an airplane is ap-proved in the aircraft industry) and license the utility only to operate the plant in a safe manner.

The topics which have been selected in the paragraphs above are considered to be only representative, but they are also considered to be among the more basic or fundamental issues the Commission might address.

6. G0ALS AND OBJECTIVES The Goals and Objectives are the essence of a Long Range Plan. However, we cannot write a comprehensive and specific set of recommendations related to Gcals and Objectives. We know little of the political pressures the Commissioners face. We have not studied the personnel problems within the Agency nor have we attempted to look at the Agency budget except that portion for research. We have restricted our discussion primarily to issues related 129

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to nuclear power plants. We have had little interaction with the Licensing

, Boards, the Appeals Board, and the legal staff. A Long Range Plan must be comprehensive and cover all aspects of the organization or it will be a failure.

! - We believe, however, that we should make our views known, including the following:

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! 6.1 We believe the Comissioners should develop a Long Range Plan as we discussed in our letter of August 13, 1985.

6.2 We believe that a Plan is a comprehensive analysis and review of 1

the activities of the Agency, the enunciation of the Comissioners' judgment of- the highest priority goals and objectives for an appropriate period of the future, and a plan of action for achiev-ing those goals and objectives. We believe the Plan should repre-sent a comprehensive, integrated course of action for the entire Agency. The Long Range Plan should be realistic, but it should also be visionary.

6.3 We wish to reemphasize the points made in our le ~ ' to Chairman Palladino on August 13, 1985 concerning the importence of a Long Range Plan for the Agency. At that time we listed the following benefits which could follow from the strategic element of long range planning;

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a. The exercise of establishing a Plan should help the Comis-sioners develop consistent Agency goals, 7

b. The existence of a Plan could lead to more deliberate regu-lation and could lend stability to the entire regulatory process.

l c. A Plan could allow a more efficient allocation of resources.  :

d. A Plan will help the Comissioners in explaining and justify-ing their programs to the Congress and the Executive Branch.

e. The existence of a Plan will help to ensure that future problems are anticipated and resolved in a timely fashion.

While the lack of a Long Range Plan does not preclude the achieve-ment of these goals, such a Plan would help ensure them.

.x l 130

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December 16, 1986 l  % ..

The Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS REPORT ON PROPOSED POLICY STATEMENT ON DEFERRED PLANTS During our 320th meeting, December 11-13, 1986, the members of the Advisory Comittee on Reactor Safeguards considered the NRC Staff's proposal for a Comission policy statement on deferred plants (see Reference). This subject was also discussed during the Safety Philosophy, Technology and Criteria Subcomittee meeting held on December 10, 1986. In our review, we had the benefit of discussions with the NRC Staff and of the document referenced.

We agree that a policy statement on deferred plants would be useful and agree in general with the Staff's recommended positions on: (1) docu-mentation, and maintenance and preservation requirements for deferred plants, and (2) procedures for reactivating deferred plants. However, it is not clear to us whether the ongoing preservation / inspection program would prevent or detect all types of equipment deterioration.

It also is not clear that the Staff would assure itself of the adequacy of a plant owner's design and construction force when reactivation begins.

We question the Staff's proposal to apply the backfit rule uniformly for deferred plants irrespective of the duration of the deferral period. We believe the proposal is too open-ended in this regard. Finally, it is not clear to us how the NRC will ensure that the owners of a deferred plant will remain knowledgeable about applicable construction and operating experience and incorporate the appropriate modifications when the plant is reactivated.

Sincerely, d

David A. Ward Chairman

Reference:

SECY-86-359, entitled " Deferred Plant Policy Statement," Heuo from Victor Stello, .Jr. for the Comissioners, dated December 2,1986.

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April 16, 1986 i

Honorable Nunzio J. Palladino -

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

SUBJECT:

ACRS COMMENTS ON QUANTIFICATION OF PUBLIC HEALTH RISKS In the course of its work, the Advisory Comittee on Reactor Safeguards has in recent months had occasion to review environmental and proba-bilistic risk assessments for several nuclear power plants. One obser-vation made as a result of these reviews .is that the NRC Staff does not normally use the best available data and health effects models in estimating the public health ri'sks to nearby population groups due to routine and potential accidental releases of radionuclides from these plants.

Because the NRC 'has been a leader in support of the development of health effects models (as represented, for example, by the publication of NUREG/CR-4214 in July 1985) and in gaining international harmoniza-tion in the application of such models (as summarized in SECY-86-8 issued on January 7,1986), we believe that the NRC Staff should be encouraged to move forward more rapidly in using these improved models in the preparation of environmental and probabilistic risk assessments for U. S. plants.

We also recommend that population dose estimates produced as a result of probabilistic risk assessments be expressed both in terms of the total dose and in terms of the number of people exposed within each dose range. In addition, we recommend that the NRC give further considera-tion to the establishment of a de minimis level for terminating the calculation of collective populatf6n doses associated with such assess-ments. Although the proposed revision of 10 CFR 20 suggests a possible cutoff at a dose rate of 1 mrem /yr, we believe that selection of such a vglue should include consideration of the total dose as well as spatial

, and temporal factors, possibly coupled with appropriate cost-benefit assessments.

Dr. Harold W. Lewis did not participate in the preparation of this report.

1 l

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Honorable Nunzio J. Palladino April 16, 1986 l

Additional coments by ACRS Member Harold W. Lewis and by ACRS Members Dade W. Moeller and David Okrent are presented below.

Sincerely, F

David A. Ward Chairman Additional Comments by ACRS Member I:arold W. Lewis It gives me great discomfort to add coments to this letter because my only real objection is to its mildness in the face of the fact that the NRC Staff has yet to apply the results of the 1980 Academy Report on the Biological Effects of Ionizing Radiation (a great improvement over the 1972 report) in a single licensing review. This is 1986 and there have been additional improvements since 1980 but they are not yet in finished form. The Comittee says "does not normally," which is, I suppose a euphemism for never.

I do wish the record to show that I did not participate in the prepara-tion of this letter. The Comittee completed and fomally approved a stronger and more complete letter at its March meeting, but it was decided during the intervening month (without benefit of a meeting) not to mail it to you. Since I believe that the final approval of a letter means just that and that this substitution is therefore improper, I could not bring myself to participate.

Additional Coments by ACRS Members Dade W. Moeller and David Okrent We believe that it is relevant to note that, at least in our opinion, the earlier letter was either inaccurate or at best unsatisfactorily incomplete, in one or more of its recomendations.

I l

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May 13, 1986 Honorable Nunzio J. Palladino l Chairman l V. S. Nuclear Regulatory Commission l Washington, D. C. 20555

SUBJECT:

SUPPORT OF RADIATION PROTECTION ORGANIZATIONS As discussed with you and the other Commissioners during our meeting on May 8,1986, the ACRS believes that the NRC should provide funds in support of the programs of certain key radiation protection organizations. These include the National Council on Radiation Protection and Measurements, the International Commission on Radiological Protection, and the National Academy of Sciences Comittee on the Biological Effects of Ionizing Radiation.

In addition to the preparation of recomendations for radiation protection standards, these organizations provide advice on subjects such as the bio-logical effects of ionizing radiation, radioactive waste management, en-vironmental radionuclide transport and surveillance, and emergency planning.

We believe that the work of these organizations is basic to the operations of the NRC and that support of these groups is one of the most cost-effective methods available to the NRC for the development of independent advice on these and related matters.

Sincerely,

.1 Y.

David A. Ward Chairman 135

e o UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION fe  :

,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

• • • • • July 16,1986 1

l 4

Honorable Lando W. Zech, Jr. 1 Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555

SUBJECT:

ADDITIONAL RECOMMENDATIONS ON THE DEVELOPMENT OF DE MINIMIS LEVELS During its 315th meeting, July 10-12, 1986, the Advisory Comittee on Reactor Safeguards gave additional consideration to procedures that might be used in the development of de minimis radiation dose values for regulatory use. This matter was also discussed during a combined meeting of our Subcomittees on Occupational and Environmental Protection Systems and on Auxiliary Systems held on June 27, 1986.

As a result of these discussions, we offer the following additional coments and recommendations:

1. Additional review and discussions need to be held to determine whether the development of de minimis values is justified. Included in such a review would be studies to determine, as far as practical, the risks (health effects) and benefits (economic savings) associated with the application of such a concept.

2. If initial studies determine that the development of de minimis dose values is justified, the development of the rationale to support the selected dose levels should be made an integral part of the overall effort.

3. To assure the acceptability of the numerical value recomended, we believe that the deliberations suggested above should be conducted by a scientific body, as contrasted to a regulatory agency. Since deter-mination of such levels will involve a wide range of considerations, including an evaluation of the risks and benefits associated with all methods for generating electricity, we believe that the organization selected to conduct these studies should have a " global" outlook. One example of such an organization is the National Academy of Sciences.

l 137 _ . _

i Panorable Lando W. Zech, Jr. July 16,1986

4. If the suggested studies lead to the development of de minimis values, careful consideration should be given as to how they should be applied within the regulatory framework.

Although the development of de minimis values is of importance to the NRC, other Federal agencies have equal interests in this matter. We suggest that the NRC consider taking the lead in joining with other appropriate Federal agencies to follow through on this issue.

Sincerely,

.I David A. Ward k.

Chairman l

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$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555

• • • • • July 16, 1986 i

Honorable Lando W. Zech, Jr.

l Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS REGARDING SUPPORT OF RADIATION PROTECTION ORGANIZATIONS SUCH AS THE NATIONAL COUNCIL ON RADIATION PROTECTION AND MEASUREMENTS (NCRP), THE INTERNATIONAL COMMISSION ON RADIOLOGICAL PROTECTION (ICRP), AND THE NATIONAL ACADEMY OF SCIENCES (NAS)

By his letter of July 1,1986, the EDO. responded to our May 13, 1986 letter to Chairman Palladino on the subject matter (copies attached).

We most certainly agree with his comment that the assessments provided by the NCRP, the ICRP, and the NAS of the significance of the results of ongoing research concerning the effects of exposures to various types of radiation delivered to various parts of the body are important to NRC's mission. Indeed, these assessments are basic and provide the most authoritative information available for the agency's decisions as to the levels of protection needed over a wide range of circumstances.

In addition, it should be noted that the range of activities of the NCRP, for example, extends far beyond assessments of the biological effects of ionizing radiation. Other NCRP program areas include the movement, transport, and behavior of radionuclides within the environ-ment, procedures for environmental radiation surveillance, nuclear emergency planning, radionuclide contamination limits, and occupational and environmental radiation protection. You might recall, for example, that it was the NCRP that provided a definitive evaluation of the potential impacts of the release of krypton-85 from the containment at Three Mile Island, Unit 2, following the accident at that plant.

As the E00 also pointed out, in the past, the NRC has supported these

! efforts to assimilate and codify the most recent information bearing on the effects of radiation. Purely as a matter of bookkeeping convenience

[since this activity is neither initiated nor controlled by the Office of Nuclear Regulatory Research (RES)], the funds for such support have, in the past, been included within the RES budget. But it is the agency 139

,--r e,-, , - - - - - - - , - - , . - . - , - - , - - , , - - - - , -

Honorable Lando W. Zech, Jr. July 16, 1986 i ,

! -- and not RES, nor any particular division of the agency -- which has an essentially primordial need for the information in question. To argue that, because funds. for the research activity of the agency have -

been subjected to such major budgetary reductions, there is no longer room in the RES budget to accomodate this support -- which may well be correct -- ought not lead to the decision to drop such. support; but ought, instead, to lead to efforts to find the mechanism whereby the-agency can continue to help with the development of the information it ,

, needs. Parenthetically, it is . obvious that no division of the agency can be expected to take the funds for such a broad purpose out of its

! own budget.

Obviously, also, the resources available to the agency as a whole have been curtailed. According to the. EDO, this has resulted in placing

" primary emphasis" on the inspection and enforcement programs, as well

as on other " program needs." Amongst the latter is the unquestionable

need of ascertaining the cause, sequence, and consequences -of the j Chernobyl accident, and the determination of what the implications of.

j this may be with respect to U.S. regulatory policies and practices.

1 However, these " program needs" are evidently somewhat flexible, as

] indicated by the intention expressed in SECY-06-185, " Program Plan for

! Chernobyl Accident Followup," of also inquiring into the Chernobyl geology, and the design features to accomodate seismic, tornado, and flood loads, the Soviet use of simulators, etc., etc., etc. -- items of interest, no doubt, but items having little if anything to do with the cause and progress of the accident, or with the implications for U.S.

i practice.

Thus, even in the context of a stated level of support, the " program j needs" admit of some flexibility -- in addition to the need of allowing

< for the possibility that support levels could change. In the radiation l protection area, the " primary emphasis on inspection and enforcement" l

could, of course, continue for years on the basis of existing data and j regulatory dicta; but, to the extent possible, it ought to adapt to the best current information available. The acquisition of such information

is a prerequisite to having these activities proceed on the most ra-tional basis possible.

The Committee believes that there are many areas within RES, as well as the remainder of the Connission's program, whose funding could be i reduced by the amount necessary to provide support to these organiza-tions with a relatively small loss compared to the substantial gains to 4

the Commission from the studies conducted by these groups.

In conclusion, while we think we understand the ED0's position, we 4 believe much more strongly that this position should be changed. An 1

140

Honorable Lando W. Zech, Jr. July 16, 1986 l

integral part of the agency's mission should be that of supporting efforts to obtain the information which it needs.

l Sincerely,

.I k.

David A. Ward Chairman Attachments:

1. Letter from Victor Stello, Jr., Executive Director for Operations to David A. Ward, Chairman, ACRS, regarding NRC funding for in-dependent scientific advisory groups, dated July 1,1986

(*) 2. Letter from David A. Ward, Chairman, ACRS, to N. J. Palladino, Chairman, NRC,

Subject:

Support of Radiation Protection Organiza-tions, dated May 13, 1986

(*) For Attachment #2, See page 135.

141 l

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'g UNITED STATES yj ,e g NUCLEAR REGULATORY COMMISSION 3 *\ " /. E WASHINGTON, D, C. 20555 e j g 0 1 1986 Dr. David A. Ward, Chairman Advisory Committee on Reactor Safeguards U.S. Nuclear Regulatory Commission l Washington, DC 20555

Dear Chairman Ward:

I appreciate the concern and comments of the Advisory Committee on Reactor Safeguards expressed in your letter of May 13, 1986, concerning NRC funding for independent scientific advisory groups in the field of radiation 4

protection. The National Council on Radiation Protection and Measurements, the International Commission on Radiological Protection and the National Academy of Sciences Committee on the Biological Effects of Ionizing Radiation play important roles in the development of radiation protection concepts, recommendations and information on radiation risks. These areas are, of course, important to NRC's mission, and the NRC has supported all of these groups during past years.

This is, however, a time of major budgetary reductions in NRC research efforts, particularly those concerned with radiation protection and health effects. These reductions have led to the necessity to terminate research in these areas. We are also experiencing reductions in staffing in these areas.

One of the reasons for cutting back in these areas is to ensure that adequate resources are available to support our nuclear safety research programs. The NRC, as you know, has the primary Federal responsibility for reactor safety and

' regulation; we do not have the primary Federal responsibility for formulating radiation protection standards. Because of the budget decreases, we are

! adjusting our resources to support our primary missions. In the radiation protection area, the NRC is placing primary emphasis on our inspection and enforcement programs to ensure that public health and safety are protected.

Consequently, although we believe that the work of these independent advisory groups is very important, we cannot continue to fund these groups while, at the same time, research and technical support programs in areas which are more directly responsive to our program needs are being cut back or terminated.

I hope that you will understand our position.

, Sincerely, Victor Stello, J[.

Executive Director for Operations ATTACTIENT l 142

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January 13, 1986 l

l Mr. Victor Stello Acting Executive Director for Operations U.S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

ACRS ACTION ON THE PROPOSED REVISION 3 TO REGULATORY GUIDE 1.63 During its 309th meeting, January 9-11, 1986, the ACRS heard a report of its Subcommittee on Regulatory Activities regarding the proposed Revi-sion 3 to Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants."

We concur in the NRC Staff's proposal to issue this Guide for public comment. Subsequent to 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 responses to them.

Sincerely, David A. Ward Chairman

Reference:

Proposed Revision 3 (Draft 4) to Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants," dated December 6, 1985.

cc: S. J. Chilk, SECY J. W. Roe, ED0 R. B. Minogue, RES G. A. Arlotto, RES S. K. Aggarwal, RES C. Bartlett, RES R. Hernan, NRR 143

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8 n NUCLEAR REGULATORY COMMISSION U <E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS "o WASHINGTON, D. C. 20555 August 12, 1986 l

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Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS ACTION ON REGULATORY GUIDE 1.114, REVISION 2, " GUIDANCE TO OPERATORS AT THE CONTROLS AND TO SENIOR OPERATORS IN THE CONTROL ROOM OF A NUCLEAR POWER UNIT"

Dear Mr. Stello:

During its 316th meeting, August 7-9, 1986, the Advisory Comittee on Reactor Safeguards concurred in the regulatory position of Regulatory Guide 1.114, Revision 2, dated June 2, 1986 and in the conforming revisions to the Stan-dard Review Plan Section 13.1.2.

Sincerely, O

t

.\ k David A. Ward Chairman 1

145

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• W ASHINGTON, D. C. 20555 e

o.,,,, December 16, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS ACTION ON THE PROPOSED REVISION 3 TO REGULATORY GUIDE 1.63, " ELECTRIC PENETRATION ASSEMBLIES IN CONTAINMENT STRUCTURES FOR NUCLEAR POWER PLANTS" During the 320th meeting of the Advisory Committee on Reactor Safe-guards, December 11-13, 1986, the Committee concurred in the regulatory position proposed in Revision 3 of Regulatory Guide 1.63, " Electric Penetration Assemblies in Containment Structures for Nuclear Power Plants."

Sincerely, 4

.l David A. Ward Chainnan

Reference:

Proposed Revision 3 to Regulatory Guide 1.63 (June 1986 version) trans-mitted to the ACRS by Memorandum from G. A. Arlotto to R. F. Fraley, dated October 14, 1986.

cc: S. J. Chilk, SECY J. W. Roe, ED0 E. Beckjord, RES G. A. Arlotto, RES S. K. Aggarwal, RES C. Ba.-tlett, RES R. Hernan, NRR 147

pa atoug UNITED STATES

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o,'n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

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• • ..+

December 16, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON PROPOSED NRC FINAL REGULATORY GUIDE ENTITLED,

" FORMAT AND CONTENT OF PLANT-SPECIFIC PRESSURIZED THERMAL SH0CK SAFETY ANALYSIS REPORTS FOR PRESSURIZED WATER REACTORS,"

DATED JUNE 1986 During our 320th meeting, December 11-13, 1986, the . members of the Advisory Committee on Reactor Safeguards heard reports from our Subcom-mittee on Metal Components and from the NRC Staff on the final draft of a regulatory guide implementing the recently adopted rule (10 CFR Part 50.61) concerning reactor vessel pressurized thermal shock (PTS) for pressurized water reactors. This matter also was considered by our Metal Components Subcommittee on February 27-28, 1986.

As a result of these discussions, we recomend that you issue the final PTS regulatory guide. However, we would like to review the work performed by the lead licensee using this regulatory guide, when it becomes available.

Sincerely.

David A. Ward Chairman cc:

E. Beckford, RES 149

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! o NUCLEAR REGULATORY COMMISSION 5 E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

• WASHINGTON, D. C. 20555 e

• • • .. March 19, 1986 l

Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON PROPOSED SAFETY G0AL POLICY During its 311th meeting, March 13-15, 1986, the Advisory Committee on Reactor Safeguards met with representatives of the Office of the Executive Director for Operations and continued its reviewThe of the issues associated Committee last reported with the issuance of the NRC Safety Goal Policy.

on this matter in a letter to you dated July 17, 1985. Since July 1985, Sub-committee meetings were held on this subject on November 6,1985, December 4, 1985, and March 12, 1986. This matter was also discussed during the Cecember 5-7, 1985 ACRS meeting. The memorandum from V. Stello to the Cemission, dated February 14, 1986, on Safety Goal Policy served as a focal point for this part of our review. We also had the benefit of comments from several ACRS consultants.

A large number of complex issues, both general and specific, need to be ad-dressed in the development of a safety goal policy. In this repert, we will comment on only some aspects of the overall subject. Our coments are listed below:

1. We favor early Commission action to adopt a form of scfety goal policy.

However, we have a number of coments concerning the particulars of the safety goal policy currently being proposed.

2. The safety goal policy should include the two qual'.tative goals of the general form recommended by the EDO. However, we are divided on whether the second qualitative safety goal should be modified to say that the societal risks to life and health from nuclear power plant operation should be less than the risks of generating electricity by viable competing technologies (rather than " comparable to or less than," as stated in the 1983 Safety Goal Policy).

3. Further, the safety goal policy statement should include explicitly the two quantitative health effect objectives. We disagree with the NRC Staff that these quantitative objectives should not appear as discrete statements of expectation in the policy statement.

151

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Honorable Nunzio J. Palladino March 19, 1986 l

4. We do not recommend including in the policy a separate statement to the effect that there should be a goal for making the likelihood of core melt small in the population of existing reactors during their lifetime.

5. We have considerable concern that quantitative cost-benefit analysis will become a major factor, if not the major factor, in decision making on safety issues, rather than being treated as only one attribute of the judgmental process.

6. We fail to see appropriate guidance for including uncertainties in decision making. The calculation of best estimate rather than con-servative values, together with a full display of uncertainties, as-sumptions, and omissions, will be in the right direction. How to factor this highly uncertain information into specific decisions remains to be deterinined.

7. We are opposed to the use of the quantitative matrix defined in the memorandum dated February 14, 1986 by V. Stello. Some members favor its use on a trial basis for a few years as one attribute ~in decision making. However, others think it is too complex. Some object to it because of inclusion of averted on-site costs. Still others think the matrix idea should be replaced by a cost-benefit approach which includes all costs. All agree, however, that cost-benefit should be only one of the several inputs into decision making.

8. In the ACRS report of July 17, 1985, the Committee said, "We believe that the Commipsion should state that a mean-core-melt frequency of not more than 10-'+ per reactor year is an NRC objective for all but a few small, existing nuclear plants, and that, keeping in mind the consider-able uncertainties, prudence and judgment will cend to take priority over benefit-cost analysis in working toward this goal." We reiterate this position.

9. In a severe accident, it is the releases from containment which consti-tute the risk to the health and safety of the public. Thus, risk cannot be assessed without a judgment on containment performance. We reiterate our recommendations to develop a containment performance objective.

10. There is a need for both the Comittee and the NRC Staff to develop guidance for implementation of the safety goals. Since plant main-tenance, operations, and management will have a significant -impact on risk, this effort should include the development of " indicators of performance" through which the acceptability of such activities can be judged.

We expect to develop additional recommendations regarding this matter during our April meeting.

152

March 19, 1986 Honorable Nunzio J. Palladino Additional coments by ACRS Members William Kerr, J. Carson Mark, and by '

Harold W. Lewis are presented below.

l Sincerely, l

David A. Ward Chairman Additional Comments by ACRS Members William Kerr and J. Carson Mark We do not favor publication of guidance which implies that it is possible to demonstrate compliance with a goal of risk for prompt fatalities of 0.1 percent of the sum of prompt fatality risks from other accidents, or of risk from cancer fatality of 0.1 percent of the sum of cancer fatality risks from other causes.

Additional Comments by ACRS Member Harold W. Lewis While I believe that there has been improvement in the drafts of a proposed Commission safety goal, I also believe that the current version remains so flawed as to not warrant issuance. Many of my reasons were detailed in my additional coments to the ACRS letters of June 9,1982 and July 17, 1985, which I believe to still be valid, and I will only add a few remarks here.

Nothing I say should be construed as opposition to the issuance of a safety goal -- I seek only higher quality.

I believe the proposed qualitative goal relating to the relative risks of other means of generating electricity is unsound, because it is based en-tirely on risk, and there is more at stake than risk. The American people might well prefer a slightly riskier alternative (if it existed) which was both free and kind to the environment.

I do not believe that there should be an individual goal, but that any few individuals who might have to bear a burden, in the comon interest, ought to be suitably compensated.

There are too many slippery targets in the proposed goals and their implemen-tation plan. Most have forgotten the arbitrary - some might say capricious -

origin of the historic ALARA number of $1000 per man-rem, but its use per-vades the implementation plan for the safety goal. The following table is relevant:

l 153

Honorable Nunzio J. Palladino March 19, 1986 1960 1970 1980 cancer death rate * (male) 162.5 182.1 205.3 cancer death rate * (female) 136.4 144.4 163.6 value of the dollar (1967=1.0) 1.127 0.860 0.406

• death rates per 100,000 population per year Surely, the Comission cannot intend that its vigilance will be allowed to relax as the natural cancer death rate increases, yet the goal says so.

Surely, the value of a human life isn't decreasing at the rate shown, yet the Comission plans to say so. Both of these problems can be circumvented by a declarative goal, as recommended in my dissent of last year, and using these numbers to put the current meaning of the goal into context.

I do not believe that on-site costs other than radiological costs should be included, for familiar reasons.

The NRC Staff supports the requirement that no one source or sequence be much larger than the others as a contributor to risk by saying that that will help to reduce uncertainty. I am unable to understand that argument, and believe it to be incorrect.

I do not support a core-melt probability goal as a surrogate for public risk.

It is not. It could be used as a regulatory tool, but should not be a Comis-sion goal.

The Comittee itself finds the proposed matrix approach to regulation unac-ceptable. So do I.

Finally, to repeat, I support the early issuance of an arbitrary quantitative safety goal along the lines suggested earlier. It is the extra baggage carried by this proposal that I find disturbing.

Reference:

Memo dated February 14, 1986 from Victor Stello, Jr., Acting Executive Director for Operations for the Comission,

Subject:

Safety Goal P.olicy, with enclosed Sumary Paper on Safety Goals 154 l

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utg'o, UNITED STATES 8

o NUCLEAR REGULATORY COMMISSION

,E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ WASHINGTON. D. C. 20555 Q

April 15,1986 Honorable Nunzio J. Palladino Chaiman U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

ADDITIONAL ACRS COMMENTS ON PROPOSED NRC SAFETY G0AL POLI STATEMENT During its 312th meeting, April 10-12, 1986, the Advisory Comittee on Reactor Safeguards continued its review of the issues associated with the promulgation of the proposed NRC Safety Goal Policy Statement. The Comittee last commented on this matter in reports to you dated July 17, 1985 and March 19, 1986. We have reconsidered these two reports and reconfimed their content. This matter was also discussed during an ACRS meeting with the Comission on March 28, 1986. The memorandum from Victor Stello, Jr., Acting EDO, to the Comission dated February 14, 1986 on Safety Goal Policy and the Draft Policy Statement on Safety Goals dated March 18, 1986 served as a focal point for this part of our review.

Issuance of Policy As stated in our report of March 19, 1986, we favor early Comission action to adopt a safety goal policy.

Qualitative Goals As stated in our report of March 19, 1986, we believe the policy statement should include the two qualitative goals as recomended by the EDO in his February 14, 1986 memorandum. We believe that nuclear power plants present substantially less public risk and offer significant environmental advantages over the only real alternative -- coal power plants. However, whether the specific comparison to other viable electric generating technology should be included in the statement of the safety goal policy is a matter which should be directly addressed by the Comissioners.

Quantitative Goals As stated in our report of March 19, 1986, we believe the Safety Goal Policy Statement should include explicitly the two quantitative health objectives as described in the memorandum from the ED0 dated February 14, 1986.

Performance Guidelines It is evident that there is no means of observing health effects at the very low levels proposed. However, the health effects to be considered result 155 l

l

Honorable Nunzio J. Palladino April 15, 1986 only from the release of radioactive materials to the environment. The releases realized in connection with the normal operation of reactor plants are below the levels which result in effects even at the very low levels pro-posed. It is also clear that if the releases which might occur in the event of accidents are small enough, or infrequent enough, the stated objective can be met; in fact, it seems probable that most existing reactors have been constructed and equipped and are operated in such a way that they already fulfill the requirements necessary to achieve the objective.

No release to the atmosphere on a scale significant in the context being considered can be envisaged except in circumstances in which the fuel of the core is severely damaged. Even in such a case, the capability of the con-tainment (provided precisely for such a purpose) should prevent the sort of release to the atmosphere which would threaten to exceed the goals.

Consequently, conformance with the goals will have to be judged on the basis of the type and frequency of accidents, on the capability of containment, and in some cases on plant location. The focus of implementation of the policy should then be directed to these aspects of existing plants.

We believe the Comission should adopt certain plant performance guidelines as one satisfactory means to assure conformance with the safety goal ob-jectives. These guidelines should be structured so that the principle of defense-in-depth is maintained. We are undecided as to whether these per-formance guidelines should be part of the policy statement or part of an accompanying directive to the NRC Staff. It may be unimportant and more a matter of style than substance. However, we believe the Comission itself should take responsibility for these guidelines and mandate their use.

We propose that the plant performance guidelines be regarded as fully ac-ceptable surrogates for the safety goal objectives. That is, if it can be shown that the guidelines are met, then it is accepted that the safety goal is met. However, if the guidelines are not met, it might still be possible to meet the safety goals. In that case, explicit analyses to show compliance with the quantitative public health objectives would be required.

There should be two performance guidelines and consideration should be given to development of a third.

The firist guideline would be that the chance of a loss of adequate core cooling' with consequent severe core damage should be less than 10-4 per reactor-year for all but a few small reactors.

The second guideline should relate to containment performance and should be such that the chance of a very lar environment should be less than 10-ge perrelease of radioactive materials to the reactor-year.

If feasible, a third guideline should be formulated concerning operational performance.

156

April 15, 1986 Honorable Nunzio J. Palladino Implementation Plan We are not satisfied with Section B, " Guidelines for Regulatory Implementa-tion," of the March 18, 1986 draft. The safety goal should be regarded c*, a The

! figure of merit by which to judge the adequacy of the regulations.

safety goal and the performance guidelines should be used in two ways:

1) as a means for judging the acceptability of proposed changes in regu-lations (such as the resolution of USIs, etc.), and

2) in the much more difficult task of assessing whether existing plants, designed, constructed, and operated to comply with past and current regulations, adequately conform with the intent of the safety goal.

Considerable thought must be given to the second *use. The question of whether it is essential that each individual plant, classes of plants, or only the general population of plants must be shown to be in conformmce is not easily answered. Another question is whether the essential chaucter-istics of plants that cause them to meet, or not to meet, the safety goals are related to their conformance with regulations or to the inherent charac-teristics of plant designs.

The Commissioners should be involved in these decisions and should state their intent to review and approve the primary features of proposals of the NRC Staff to implement the safety goal policy. Because of the complexity of the issues and the lack of experience in application, a review should be per-formed by the Commissioners in two or three years.

The Committee recomends that the Comission include in its policy a state-ment of intent that future nuclear power plants should be designed and operated even more safely and reliably than existing plants.

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

Sincerely, f

i i , 4.3 David A. Ward Chaiman Additional comments by ACRS Member Harold W. Lewis I have nothing new to add to my earlier additional remarks, other than to say that this Committee report is closer to my own personal views than have been the previous ones. On the other hand, I have often been asked to provide a statement of just what I would like to see the Comissian issue as a Safety Goal Policy Statement, and I have therefore taken the liberty of writing a draft, written as if I were the NRC, which I am well aware is not the case.

J 157

I Honorable Nunzio J. Palladino April 15, 1986 I believe that detail beyond that contained here should not be in a Commis-sion Policy Statement, but rather in Staff documents. Obviously, one may choose different numbers from those in the sequel, but the pattern of the statement is the substance of the following.

In the Atomic Energy Act of 1954, as amended, the Congress directed that the development of the peaceful uses of atomic energy be a national objective, and later directed the Nuclear Regulatory Comission to assure that such development proceed without compromise to the public health and safety or to the comon defense and security. The former has become the major preoccupa-tion of the Comission, and we believe that the objective has so far been met. Nonetheless, just as a certain measure of risk is inevitable in the use of any technology, and in particular in any energy-producing technology, it is present for nuclear power, and the question of just how much risk is acceptable (what is commonly termed "how safe is safe enough?") must be answered. We have no guidance from the enabling legislation, and feel that we should set forth some principles, preferably quantitative, to help our Staff understand and implement our objectives, and to help the public judge them. This document is our statement of those principles, but it must be emphasized that they are just that -- principles -- and are not a new set of regulatory criteria that must be met by each licensed plant. They are rather a set of standards by which we and others may judge the effectiveness of our regulatory program.

As a cardinal principle, we intend to keep the risk from nuclear power so low that the risk of nuclear power does not play an important role in national decision making about energy choices. There are clearly other relevant factors -- environmental impact, cost, esthetics, land use, etc. -- but the limitation of risk is our responsibility.

A principle of this kind suffers from some inherent difficulties. In the first place, it is not quantitative, so it lacks specificity. It must be made quantitative before it can be judged, but then the second problem arises, which is that any quantitative assessment of risk, in the current state of the art, is beset with large uncertainty. Thus, we do not expect that a quantitative goal can be used as a threshold objective or regulatory tool, at least in the foreseeable future.

Nonetheless, it is useful to say what we mean. We mean that the probability of a large and unacceptable release of radioactivity from a nuclear power plant should be held, on the average, below one chance in a million per plant year. We will put that number in perspective below, in tenns of current risks in life in America, but it means specifically that, with our prospec-tive population of about a hundred nuclear power plants in operation, there is an overwhelming probability that such an unacceptable event will not occur in our lifetimes, or those of our children, or those of their children, and on and on for hundreds of generations. Presumably, the technology will evolve and improve over time, so that future Nuclear Regulatory Comissions will be able to do even better.

158

Honorable Nunzio J. Palladino April 15, 1986 l

l What remains is to ask how we can test whether we are achieving our objec-tive. The tool for such a determination is probabilistic risk assessment, which can help enormously in this task, but which still involves large uncertainties, particularly in the " bottom line" assessments required here.

Even so, we need to deal with the bottom line in order to measure our effec-tiveness so that it is, as a matter of expediency, necessary to break the task down into manageable pieces which are subject to analysis, still with uncertainty. We believe that we are now meeting our objective by, very roughly, the reactormaintaining)a probability building of less than a chanceofincore-melt (molten ten thousand core per year, withon a the floor chance of less than one in a hundred that such an event will lead to a major release of radioactivity outside the containment building, which is, after all, designed to provide just such protection. This division, part of our overall philosophy of defense-in-depth, is of course arbitrary and may change with time and knowledge, but this is the way we see the protection divided at this time.

It is useful to put the risk to which we have limited ourselves here into perspective, by comparing it to other risks in our society. Obviously, such a comparison cannot be precise, but it should serve to provide a measure of the way in which we rate that due to nuclear power, in familiar tenns.

Nuclear risk, since it involves the occurrence of a highly improbable ac-cident, will, of course, always be more difficult to calculate. For example, the major damage to the health and safety of the public in the event of a radiation release of the sort weInhave described is in the late somatic the United States every year approxi-effects, in other words cancer.

mately 400,000 people die of cancer from various causes. That is a number which has been going up for some cancers, notably lung cancer for known reasons, and down for others, but the age-adjusted rate has not changed dramatically in recent years. If, in the event of a truly catastrophic nuclear accident, a thousand people were to succumb to cancer over the next few decades, as a result of the radiation released, a probability of one in a million per year would mean, for the entire country, an average mortality rate of one tenth of a person per year, for a population of a hundred nuclear plants. This should be compared with the 400,000. If the population in the vicinity of the plant (those most directly affected) were, say, 100,000, this would contribute less than one one-thousandth of one percent of their cancer risk. By the same token, accidents from other causes (half of them related to motor vehicles) lead to about 100,000 imediate fatalities per year, and here too the contribution from nuclear power will be comparably low.

It is worth noting that current estimates for the rate at which coal, the only presently viable competing technology, causes fatalities are far higher.

As stated, we de not intend that these objectives be taken as new regulatory hurdles, but believe that the numbers we have chosen are reasonable. It is our intent to mr.intain this high level of safety, and as a first step, we will ask our staff to assure us, on an annual basis, that it is being 159

Honorable Nunzio J. Palladino April 15, 1986 maintained. The ED0 should provide us in the near future with a plan des-cribing how they will do that.

References:

1. Memorandum dated February 14, 1986 from Victor Stello, Jr., Acting Executive Director for Operations, to the Comission,

Subject:

" Safety Goal Policy," with enclosed Sumary Paper on Safety Goals for the Operation of Nuclear Power Plants

2. Memorandum dated March 24, 1986 from Victor Stello, Jr., Acting Execu-tive Director for Operations, to the Comission,

Subject:

" Safety Goal Response to Memorandum from S. Chilk to V. Stello dated March 6, 1986." with enclosed Draft Policy Statement dated March 18, 1986 and NRC/PUC Interaction 160

1. UNITED STATES

1. 'o

~,,

8 o NUCLEAR REGULATORY COMMISSION

{o E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20656

• • • • • November 10, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

APPLICATION OF NRC SAFETY G0ALS IN LICENSING ISSUES Although there is a range of questions to be answered before the recently published NRC Safety Goal Policy Statement can be fully interpreted, there is one step that we believe might be accomplished with minimal effort at this time. That would be to develop a system of tables that indicate the doses to the population that would be compatible with the health effects goals ex-pressed in the Policy Statement. Such a development would also provide an opportunity for formal application of the health effects models recently updated under the auspices of the NRC Staff.

Although this is but a small part of the total process, completion of this step would enable those groups conducting PRA studies to more readily inter-pret the results of their efforts in terms of the Safety Goals. Because we also believe that such action would be beneficial in furthering the use of the goals, we recommend that you consider arrangina to have the indicated tables prepared.

Sincerely, l

. D David A. Ward Chairman 1

161

f

/*  %'n UNITED STATES NUCLEAR REGULATORY COMMISSION l I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

$ WASHINGTON, D. C. 20666 s

o,... February 19, 1986 l l l

Mr. Victor Stello Acting Executive Director for  ;

Operations l U. S. Nuclear Regulatory Comission Washington, D. C. 20555

SUBJECT:

ACRS C0teENTS ON PROPOSED REVISION OF 10 CFR PART 20, i " STANDARDS FOR PROTECTION AGAINST RADIATION" During its 310'.h meeting, February 13-15, 1986, the Advisory Comittee on Reactnr Safeguards heard a report from its Subcomittees on Waste Managtment and Reactor Radiological Effects regarding the Proposed Revision of 10 CFR Part 20, " Standards for Protection Against Radiation." The Subcommittees held a joint meeting on January 15-17, 1986 during which they heard presentations by the NRC Staff and the Atemic Industrial Forum. The ACRS discussed and approved the Subcom-

~

mittees' comments which are transmitted herewith for your consideration.

Additional coments by ACRS Member Max W. Carbon are presented below.

Sincerely, j

David A. Ward Chaiman Additional Coments 'by ACRS Member Max W. Carbon:

~

I believe the proposed revisions to Part 20 are not needed and are not cost effective. I do not support the Comittee's view.

l

Attachment:

ACRS Coments on Proposed Revision of 10 CFR Part 20. " Standards for Prc tection Against Radiation" l

163 I

i.

e ,

t Mr. Victor Stello February 19, 1986 i

Reference:

. Federal Register, Part II, Nuclear Regulatory Comission, 10 CFR Parts 19, 20, et al., Standards for Protection Against Radiation; Proposed Rule; Extension of Coment Period and Republication, Vol.

51, No. 6, dated January 9,1986 cc: Chairman Palladino

, Comissioner Roberts

. Comissioner Asselstine 1

Comissioner Bernthal Comissioner Zech K. Goller, D/DRPES, RES R. Alexander, DRPES, RES l D. Harward, Atomic Industrial Forum 4

t j

I 1,

i 164

.. - . . - _ - - . a,. . . - - , - -. . - . - _ - - - _ _ . - _ _ _ .

(

t*

s ACRS COMMENTS ON PROPOSED REVISION

?^, OF 10 CFR PART 20. " STANDARDS FOR PROTECTION l AGAINST RADIATION" i .

General Comments:

1. Th*! costs of-implementing the program will have widely variable 4

impacts on the using institutions; in a utility it will be lost in the " noise" of operational costs. On the other hand, small " users" could find the changes a significant burden. In these cases, the-inevitable economic impact of broad general regulations applicable gt to everybody should be recognized and allowances made to make the

-transition reasonable. We concur in the proposed five-year target date for final implementation of the Standards. This'should avoid ~

any unnecessary burden on most licensees.

2. We recommend that the NRC encourage and assist in the development 4

of training programs to assist licensees in understanding and implerenting the revised Standards. We consider such programs to

be a mandatory part' of the irrietantation procedure.

3. Although the main goal in preparing this revision is to provide scientific updating of-the standards, we believe that greater empha' sis"should be placed on the associated improvements that are anticipated in terms of occupational and environmental protection.

These include reductions in the annual whole body dose limit to 5 rem, increased emphasis on the implementation of the ALARA 165

Comments / Standards for Protection Against Radiation criterion, recording of more accurate and useful data on internal and external exposures, sunnation of doses from internal and external sources, reduction in the dose limit for extremities, and corrections (reductions) in the annual intake limits for alpha emitting radionuclides.

4. We believe the revision provides an excellent opportunity to move the U. S. radiation protection community into the everyday use of the International System of Units (SI). This aspect of the Stan-dards should be strengthened. The nomenclature used in the report i should also be made compatible with the existing scientific data base. At present, the revision perpetuates the use of obsolete terms, such as " absorbed dose," and provides definitions and applications of terms and concepts that are inconsistent with current (ICRP) usage.

Specific Comments:

1. The proposed approach for protecting the fetus (Section XII) appears nebulously worded. We recommend that more positive action be considered.

2. The NRC Staff has suggested (Section XVIII) that individual dose rates of 1 mrem per year or less be considered below the limit of 166

Coments/ Standards for Protection . Against Radiation regulatory concern and that this dose rate be used as a cutoff for calculating collective doses to the population. Although we comend the staff for this action, the current text seems to have been designed to support a value of 0.1 mrem and inadequately justifies the choice of 1 mrem. In addition, Figure 1 (Section XVIII) is not clear on the dose rate that would be considered de I

minimis for the most exposed individual member of the public. This should be clarified.

3. In connection with collective dose calculations, we suggest that data on population dose include not only the collective dose but also the number of people within each dose range. Although the collective dose is important, we believe the presentation of the additional data, as described, will be useful in providing infor-

. mation on the number of people within each range of risk.

4. We see no justification (Section XXV) for exempting excreta from

medical patients from regulatory control. The rationale as pre-sented is not adequate. Most importantly, we can see no justifi-cation for exempting exposures from such discharges from being included in dose assessments for individual members of the pu'blic l (SECY-85-147).

{

5. Section XXIII refers to " Disposal into Sewerage." The term " sewer-l age" refers to the sewers and the sewage they contain. This Section would more properly be titled, " Disposal into Sewers."

167 t

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Coninents/ Standards for Protection Against Radiation

6. A distinction should be made in the words used on warning signs which designate radiation areas in contrast to high radiation areas. In our opinion, the word " danger" should be reserved for signs designating high (and very high) radiation areas; the word

" caution" should be used on signs designating radiation areas in general. -

7. In considering several of the Regulatory Guides that provide supporting information for radiation protection programs, we note that the NRC Staff considers that certification in the nuclear power plant specialty by the American Board of Health Physics (ABHP) is adequate confirmation of the qualifications of a person to serve as a Radiation Protection Manager at a commercial nuclear power plant. We believe that certification by the ABHP in the comprehensive practice of health physics should be equally con-sidered by the NRC Staff as adequate confirmation of the quali-fications of a person to serve as a Radiation Safety Officer or Manager for other licensees such as a university, a hospital, or a major research or industrial installation.

I 168

'o UNITED STATES g

e n NUCLEAR REGULATORY COMMISSION 4 E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASMNGTON, D. C. 20555

• • • • • March 19, 1986 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON PROPOSED BROAD SCOPE RULE' REVISION TO GENERAL DESIGN CRITERION 4 During its 311th meeting, March 11-13, 1986, the Advisory Committee on Reactor Safeguards met with representatives of the NRC Staff and reviewed the proposed changes to General Design Criterion 4 (GDC 4).

The ACRS Subcomittee on Metal Components held a meeting on this subject on February 27-28, 1986 with representatives of the NRC Staff and the nuclear industry. We also had the benefit of the document referenced.

While the assumption of an instantaneous double ended pipe break in large high energy lines has provided a convenient way to bound the loads that might result from a pipe rupture, it bears little or no relationship to the way in which such pipes have actually developed leaks, and has led the Comission to require massive structures to protect against the consequences of these hypothetical breaks. It has also led to systems which are expensive to build, difficult to in-spect, and arguably more prone to failure than those without the large restraints.

In looking for a more rational yet safe approach to pipe breaks, the research of the last decade on elastic-plastic fracture mechanics has presented a means to justify a more reasonable alternative. This work had shown how to calculate when a large through-wall crack in a high energy line will be stable instead of leading to sudden failure; that is, when " leak-before-break" will occur. The formalism has been well proven as a means for determining how fast a crack will grow and when these cracks will become unstable. In addition to work in the nuclear industry, this technology has been applied for many years in aircraft structures and in large structures like off-shore drilling platfonns.

The proposed revision of GDC 4 would allow the use of leak-before-break methodology to exclude dynamic effects associated with postula-ted pipe ruptures. We believe that this is an appropriate approach and that the proposed rule change should be allowed to go out for public commonts. Subsequent to the public comment period, we expect to review this rule together with the public comments and the NRC Staff's response to them. We also wish to be kept informed of the detailed acceptance cri teria developed by the NRC Staff for the 169

1 Honorable Nunzio J. Palladino March 19,1986 selection of piping systems to which this revision will be applied.

We give below some comments on this proposed revision.

The proposed change to GDC 4 would add two sentences, as follows:

"However, dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping.

These analyses must include, as a minimum, a deterministic fracture mechanics evaluation of the piping, and an evalu-ation of corrosion, water hammer, fatigue, leakage de-tection, and indirect sources of pipe rupture."

We recommend that the second (last) sentence be deleted. The require-ment in the first sentence that "... analysis demonstrate that the probability of fluid system pipe rupture is extremely low ..." covers all that should be said in a general rule. To attempt to spell out detailed requirements in this way:

1eads to an incomplete list, for example, creep, creep-fatigue, and erosion are credible but are omitted, brings into question whether the rule would apply to some ad-ditional failure mechanism that might be found to be important in the future, and 1eads to possible contradictions such as requiring a " deter-ministic fracture mechanics analysis" on a pipe with an "ex-tremely low probability of rupture."

The rule should not be applied in the foreseeable future to piping operating at temperatures above 750 F. The criteria that the NRC Staff is considering to evaluate piping deals primarily with the development of fatigue cracks. This mechanism of failure gives a clearly defined crack which grows slowly with time and is detectable with the NDE techniques used. At higher temperatures creep damage causes more homogeneous deterioration of the metal. This deterio-ration is not detectable with the nondestructive examination tech-niques commonly used by the industry and can lead to sudden failures which give little warning. It is believed that the recent sudden failures in fossil plant steam lines operating at 1000 F were of this type.

T'ie text introducing the rule change brings up other concerns, the primary one being that in attempting to formulate " rigorous acceptance criteria" the NRC Staff may impose requirements at least as onerous as those now required. Though rigor has its place, judgment might be more appropriate in the evaluation of potential problems with corro-sion, water hammer, indirect sources of pipe rupture, etc. In a 170

Honorable Nunzio J. Palladino March 19, 1986 similar vein, the package of materials properties for the various joints could balloon out of all proportions.

Additional coments by ACRS Members David Okrent and Glenn A. Reed and ACRS Members Paul G. Shewmon and David A. Ward are presented below.

Sincerely, ii David A. Ward Chairman Additional Coments by ACRS Members David Okrent and Glenn A. Reed We disagree'with the proposed modification at this time. Although we can see the economic and safety benefits from removal of pipe re-straints, we do not believe that the NRC Staff has made a case for going beyond what the regulatory authorities currently allow for PWRs in the Federal Republic of Germany (FRG) nor have they defined the detailed conditions of design, fabrication, inspection, and monitoring for reactors in operation or under construction that would be equiva-lent to the requirements imposed in the FRG, where the FRG permits use of leak-before-break.

In addition, it appears that the proposed modification of GDC 4 might encourage licensees with nonqualifying situations to pursue relaxation and might permit undesirable relaxation in subcompartment pressuri-zation capabilities as well as undesirable routing of high energy lines.

We suggest that the NRC Staff be asked to resubmit a revised broad scope approach.

Additional Coments by ACRS Members Paul G. Shewmon and David A. Ward We believe it is illogical to apply an argument to one aspect of nuclear power plant design without applying it to other aspects where it equally applies. Therefore, we believe the NRC Staff should begin to consider whether leak-before-break analysis could also be usefully applied to questions of emergency core cooling, containment design, and environmental equipment qualification with an accompanying poten-tial for reduction in both risk and restrictive regulations.

Reference:

1. Letter from G. A. Arlotto, Office of Nuclear Regulatory Research, to R. F. Fraley, Advisory Committee on Reactor Safeguards,

Subject:

Updated Broad Scope GDC 4 Rule, dated February 9,1986 171

pa Mrug'o UNITED STATES g

/ o NUCLEAR REGULATORY COMMISSION

$ ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS e WASHINGTON, D. C. 20556

%*****/- July 15,1986 l

Mr. Victor Stelio, Jr.

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

SUBJECT:

ACRS ACTION ON THE PROPOSED REVISIONS TO SECTIONS 9.2.1 AND 9.2.2 0F THE STANDARD REVIEW PLAN (SRP)

During its 315th meeting, July 10-12, 1986, the Advisory Committee on Reactor Safeguards heard a report of its Subcommittee on Auxiliary Systems regarding the proposed revisions to the SRP Sections 9.2.1,

" Station Service Water System," and 9.2.2, " Reactor Auxiliary Cooling Water Systems." These revisions incorporate the resolution of Generic Issue 36, " Loss of Service Water (Calvert Cliffs, Unit 1)" into the SRP.

The ACRS accepted the recommendation of its Subcommittee that these revisions were acceptable.

Sincerely, cu RahondF.Fraley Executive Director

Reference:

Proposed Revision 4 to SRP Section 9.2.1 and Revision 3 to SRP Section

9.2.2 dated June 1986, transmitted by a memorandum from H. Denton, NRR, to J. Sniezek, CRGR, and R. F. Fraley, ACRS, dated May 13, 1986.

,cc : S. J. Chilk, SECY J. W. Roe, EDO R. Hernan, NRR I

173

/ %o UNITED STATES

! NUCLEAR REGULATORY COMMISSION

{o I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS .

wasumorow, o.c.aosse September 16, 1986 Honorable Lando W. Zech, Jr.

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

SUBJECT:

ACRS COMMENTS ON THE PROPOSED REVISION TO THE ECCS RULE IN 10 CFR 50.46, " ACCEPTANCE CRITERIA FOR ECCS FOR LIGHT WATER NUCLEAR POWER REACTORS," AND APPENDIX K, "ECCS EVALUATION MODELS" During its 317th meeting, September 11-13, 1986, the Advisory Comittee on Reactor Safeguards reviewed the NRC Staff proposal to issue for public coment a revision to the ECCS rule contained in 10 CFR 50.46 and Appendix K. In its review, the Committee had the benefit of discussions-with representatives of the Office of Nuclear Regulatory Research (the sponsors of the rule revision) and the Office of Nuclear Reactor Regu-lation. Subcomittee meetings on this topic were held on April 29-30 and August 28, 1986. The Committee also had the benefit of the docu-ments referenced.

There is a long history of emergency core cooling issues and evolution of the ECCS rule which significantly predates proposed issuance of this l rule revision. The ACRS has had a long-standing involvement with these issues and, over the years, the Committee has urged that the rule be carefully evaluated for possible revision in recognition of LOCA re-search which demonstrated substantial core cooling margins. The NRC Staff had a revision effort underway in 1978-79, but this effort was '

curtailed by the TMI-2 accident and resulting regulatory actions. In 1983, the NRC endorsed an ' interim revision approach which encouraged improved evaluation models based on realistic calculations combined with an uncertainty evaluation to demonstrate an adequate margin of ' safety.

Since this approach was tied to the existing ECCS rule requirements, the extent of model improvement was limited.

The approach now proposed by the NRC Staff complements the interim revision noted above. The revised rule would eliminate the requirement to use the models specified in- Appendix K and allow use of realistic models combined with an uncertainty analysis of the overall calculation.

Certain 10 CFR 50.46 limits, such as 2200"F' peak clad temperature and 17% cladding oxidation, would be maintained as bounds on the calcula-tion. The current Appendix K requirements would also be grandfathered indefinitely for licensees who elect to use the present evaluation model 175

Honorable Lando W. Zech, Jr. September 16, 1986 approach. A proposed regulatory guide accompanies the revised rule.

While we support the intent of the revised rule, we offer the following coments:

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

• The proposed regulatory guide lacks sufficiently detailed guidance, particularly in the areas of uncertainty calculations and those ,

features of the models that would be acceptable to the NRC Staff.  !

The guide should be indexed so that it corresponds more closely to the general provisions of the rule.

• A Compendium of ECCS research will be issued in support of the rule. We understand that RES plans to submit the compendium to peer review. We wish to review the final document.

We are not yet convinced that the current ECCS rule should be grandfathered indefinitely.

We believe that the revised ECCS rule, and the associated regulatory guide, should be issued for public coment. The Comittee expects to complete its review of the proposed rule package after NRC Staff con-sideration of the public coments.

Sincerely, David A. Ward Chairman

References:

1. U.S. Nuclear Regulatory Commission Draft SECY paper for the Comis-sioners from V. Stello, EDO, " Revision to the ECCS Rule Contained in Appendix K and Section 50.46 of 10 CFR Part 50," provided to the ACRS August 18, 1986

2. U.S. Nuclear Regulatory Commission Draft NUREG Document, "Compen-dium of ECCS Research for Realistic LOCA Analysis," Office of Nuclear Regulatory Research, transmitted by memorandum to ACRS from ,

L. Shotkin, NRC, dated August 12, 1986 j 176

(

/ 'o,, UNITED STATES NUCLEAR REGULATORY COMMISSION

! n

{o ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wAsmNGTON, D. C. 20555 October 15, 1986 Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON PROPOSED REVISIONS TO STANDARD REVIEW PLAN SECTIONS 6.5.2, " CONTAINMENT SPRAY AS A FISSION PRODUCT CLEANUP SYSTEM," AND 6.5.3, " FISSION PRODUCT CONTROL SYSTEMS AND STRUCTURES" The combined ACRS Subcommittees on Severe (Class 9) Accidents and Nuclear Plant Chemistry met on September 24, 1986 to discuss portions of the NRR Implementation Plan for the Severe Accident Policy Statement, including proposed revisions to SRP sections 6.5.2, " Containment Spray as a Fission Product Cleanup System," and 6.5.3, " Fission Product Control Systems and Structures." These matters were also discussed by the Committee during its 318th meeting, October 9-11, 1986.

As a result of these discussions, we concur with the general approach being proposed by the NRC Staff with respect to PWR containment spray systems. While we also support the proposed approach being taken for BWRs, we believe that the increased dependence on suppression pools does not justify the relaxation of existing requirements relative to standby gas treatment systems.

We would appreciate an opportunity to review these two SRP sections again prior to their issuance in final form.

Sincerely,

. k David A. Ward Chairman

Reference:

i Revision to Standard Review Plan Sections 6.5.2 (Draf t), " Containment i Spray as a Fission Product Cleanup System," and 6.5.3 (Draft), " Fission 177

Mr. Victor Stello, Jr. October 15, 1986 1

Product Control Systems and Structures," received by ACRS on September l 9, 1986 1

CC: i Chairman Zech Comissioner Roberts Comissioner Asselstine Comissioner Bernthal Comissioner Carr S. Chilk, SECY Z. R. Rosztoczy, NRR 1

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178

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1. g 8 o NUCLEAR REGULATORY COMMISSION l {o I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS wasumaTow.o. c.aoses November 12, 1986

Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COM ENTS ON PROPOSED REVISED STANDARD REVIEW PLAN SECTION 3.6.2, " DETERMINATION OF RUPTURE LOCATIONS AND DYNAMIC EFFECTS ASSOCIATED WITH THE POSTULATED RUPTURE OF PIPING" DATED OCTOBER 2, 1986 The combined ACRS Subcommittees on Metal Components and on Structural Engi-neering' met on May 23 and 24,1985 to discuss the NRC proposed revisions to Standard Review Plan (SRP) Section 3.6.2, " Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping." In addition, this matter has been discussed during review of specific licensing applications where exceptions to the present rule on arbitrary intermediate pipe break locations were approved. During its 319th meeting, November 6-8, 1986, the ACRS discussed the revised SRP requirements and their application.

As a result of these discussions, we concur with the proposed revision to SRP Section 3.6.2 as it relates to modifying pipe break sizes and locations involving a need for pipe restraints and jet shields. However, we believe that the existing SRP requirements for postulating break sizes and locations should be retained where they relate to establishing compartment and subcompartment pressure buildup, particularly outside the primary containment.

Sincerely,

. Q.J David A. Ward Chairman 179

l nog \ UNITED STATES {

f .o NUCLEAR REGULATORY COMMISSION I

l  ;! ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 1 wAmaworow.o.c. zones

..... l December 17, 1986 Mr. Victor Stello, Jr.

Executive Director for Operations U. S. Nuclear Regulatory Commission

! Washington, DC 20555 4

Dear Mr. Stello:

4

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SUBJECT:

ACRS COMMENTS ON THE INTERPRETATION OF 10 CFR PART 50, GENERAL 1 DESIGN CRITERION 4 " ENVIRONMENTAL AND MISSILE DESIGN BASES" It has come to our attention that the NRC Staff is interpreting GDC 4 to mean that the capabilities of component supports can be substantially reduced in existing plants as a consequence of the elimination of the

requirement to consider dynamic effects of pipe ruptures (Reference 1).

, As you may know, NRC-funded research at Lawrence Livermore National

! Laboratory on the seismic risks to piping failure had as a principal l

conclusion that the failure of component supports constituted the largest single risk of pipe failure. The proposed Rule, " Modification of General i Design Criterion 4 Requirements for Protection Against Dynamic Effects of

Postulated Pipe Ruptures," (Reference 2), under " Scope of Rulemaking,"

1 states, "The use of ASME code allowables in component support

! redesign is judged sufficient for preventing pipe rupture due to component support failure."

However, it goes on to say,

" Redesigned component supports must have sufficient margins such that component support failure is a

- remote cause of pipe rupture."

! It is the opinion of two of the NRC's principal consultants in this area j that the ASME Code allowable values do not provide sufficient margin to

! make such failure a remote cause of pipe rupture.

We are concerned about this difference of opinion regarding the required level of safety, and we reconenend that you stop approving such requests for the reduction in supports until there has been an opportunity for i these differences to be resolved.

i 181

Mr. Victor Stello, Jr. December 17, 1986 We will be happy to work with you on the resolution of this matter. We plan to meet with the NRC Staff on this and other related matters on January 15 and/or 16,1987.

Sincerely, M e k David A. Ward Chairman

References:

1. Safety Evaluation by the Office of Nuclear Reactor Regulation, Supporting Amendment No. 89 to Facility Operating License No. DPR-72, Florida Power Corp., et al. Crystal River Unit No. 3 Nuclear Generating Plant, Docket No. 50-302, transmitted May 23, 1986

2. Federal Register Notice, Proposed Rule, " Modification of General Design Criterion 4 Requirements for Protection Against Dynamic Effects of Postulated Pipe Ruptures," Vol. 51, No.141,~ July 23,1986 182

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~g 8 n NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS o wAsmworow, o. c.zosos

• • • • • February 19, 1986 l

Honorable Nunzio J. Palladino Chainnan

, U. S. Nuclear Regulatory Commission l Washington, D. C. 20555 i

SUBJECT:

ACRS COMMENTS ON PROPOSED INSIDER SAFEGUARDS PROVISIONS During its 310th meeting, February 13-15, 1986, the Advisory Comittee on Reactor Safeguards was briefed by the NRC Staff on the subject of the Insider Safeguards Rules proposed by the Staff, and by industry representatives

(NUMARC, in particular), on the Guidelines, which utility executives were said to prefer, for achieving a similar purpose. The general subject of Insider Safeguards' provisions was also discussed during our 281st meeting, August 31-September 1,1983, and reported on in our letter of September 6, 1983. Some of these matters were reviewed again in a meeting of the ACRS Subcommittee on Safeguards and Security on January 7, 1986.

Three separate, though related, items are included in the Staff proposed Insider Safeguards Rules: the item concerning Access Authorization, the item concerning Search Requirements, and some Miscellaneous Safeguards-Related Amendments. In our letter of September 6, 1983, we generally approved of the .

provisions being considered; these provisions have not changed greatly in the 4 meantime. The following coments are addressed only to the Access Authori-l zation Item.

Recently a so-called "NUMARC initiative" has been developed; it is the consensus of the nuclear utilities that the proposed NUMARC Guidelines are a preferable alternative to the issuance of a rule by the Comission. It appears that the sort of screening required in connection with an individual being authorized for unescorted access to vital areas of a plant would be very similar under the two proposals. The major differences brought to our attention were (1) the specific provision for an " appeals process" provided in the Staff's Rule to be available in the event that authorization for unescorted access were refused, and (2) the question of whether and how compliance should be ensured in the absence of a rule. Subject to resolution of these differences, the Staff has acknowledged that the industry proposal represents an acceptable option.

! With respect to the first of these, we were informed (but have not been able

, to confirm) that an appeals context of a federally (NRC) process would be a necessary provision in theimpose i 183 i

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Honorable Nunzio J. Palladino February 19, 1986 practices voluntarily adopted by private employers. In the latter case, all existing regulations providing protection against discrimination, violation of civil rights, and so forth, would continue to apply. In addition, persons currently employed in the security forces of the utilities go through essen-tially the same screening, and no special appeals process is considered necessary on that account; some licensees already follow similar practices with other than security force employees.

With respect to the second of the above matters (ensuring compliance), on the assumption that the utilities commit to adding the screening procedures to their security plans, we suppose it would be possible to inspect for com- l t

pliance and proceed much as in the case of any other licensee commitment.

In addition, the specific matter of "grandfathering" access authorization would appear to deserve further consideration and definition by the utilities and the Staff.

In our opinion, the industry-proposed approach is to be preferred, at least on a trial basis, with the option of following later with a rule in the event that should seem necessary. This process could be initiated on the basis of a Policy Statement issued by the Comission.

Sincerely yours.

O ej k, David A. Ward Chairman 184

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UN:TED STATES NUCLEAR REGULATORY COMMISSION D E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS 4 wAssiworow. o. c. rosss k February 19, 1986 The Honorable Thomas P. O'Neill, Jr.

l Speaker of the United States House of Representatives Washington, D.C. 20510

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 (ACRS) submits herewith its comments on the Nuclear Regulatory Comission's (NRC's) Safety Research Program for Fiscal Year 1987 We note with increasing concern the continued decrease in the level of funding available for the NRC's safety research program. We are con-vinced that a continuing research program is needed in order for the NRC to fulfill its regulatory responsibilities effectively and fairly. We believe that there is some level of funding below which a research program will be ineffective or impractical, but do not now know what that level should be. The NRC has similar concerns, and has comis-sioned the Comittee on Safety Research uf the National Research Council to undertake a study and make recomendations regarding the NRC's future safety research activities. Although the NRC certainly will inform the Congress of the recomendations of that Comittee, we believe that our charter, as well as the request from the Congress for us to review the safety research program, suggests, or requires, that we review that report and provide the Congress with an independent evaluation of its recomendations together with our view on the content and appropriate funding level for an NRC safety research program in the future. We propose therefore to provide you with such a report within about six months after the report of the National Research Council Comittee has been received.

At this time, we request your recorsideration of the statutory require-ment that we provide the Congress each year with a report on the NRC's proposed research program and budget. We believe that it would be more useful to the Congress if we provided comments to you on the research

! program from time to time as seems appropriate to the issues.

Sincerely, S , &

David A. Ward i

Chairman 185

The Honorable Thomas P. O'Neill February 19, 1986 Attachments:

1. Advisory Committee on Reactor Safeguards, U.S. Nuclear Regulatory Commission, " Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Year 1987 - A Report to the Congress of the United States of America." dated February 1986.

2. Letter from David A. Ward, Chairman, Advisory Committee on Reactor Safeguards, to Nunzio J. Palladino, Chairman, U.S. Nuclear Regula-tory Commission,

Subject:

ACRS Comments on the NRC Safety Research Program and Budget for Fiscal Year 1987," dated June 11, 1985.. ,

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~g UNITED STATES F o NUCLEAR REGULATORY COMMISSION y ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 g #g

% February 19, 1986 1

l The Honorable George H. W. Bush President of the Senate Washington, D.C. 20510

Dear Mr. President:

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 (ACRS) submits herewith its comments on the Nuclear Regulatory Commission's (NRC's) Safety Research Program for Fiscal Year 1987.

We note with increasing concern the continued decrease in the level of funding available for the NRC's safety research program. We are con-vinced that a continuing research program is needed in order for the NRC to fulfill its regulatory responsibilities effectively and fairly. We believe that there is some level of funding below which a research program will be ineffective or impractical, but do not now know what that level should be. The NRC has similar concerns, and has commis-sioned the Committee on Safety Research of the National Research Council to undertake a study and make recommendations regarding the NRC's future safety research activities. Although the NRC certainly will infonn the Congress of the recommendations of that Committee, we believe that our charter, as well as the request from the Congress for us to review the safety research program, suggests, or requires, that we review that report and provide the Congress with an independent evaluation of its recomcendations together with our view on the content and appropriate funding level for an NRC safety research program in the future. We propose therefore to provide you with such a report within about six months after the report of the National Research Council Committee has been received.

At this time, we request your reconsideration of the statutory require-ment that we provide the Congress each year with a report on the NRC's proposed research program and budget. We believe that it would be more useful to the Congress if we provided comments to you on the research program from time to time as seems appropriate to the issues.

Sincerely, e k.

David A. Ward Chairman 187

l The Honorable George H. W. Bush February 19, 1986 Attachments:

1. Advisory Committee on Reactor Safeguards, U.S. Nuclear Regulatory Commission, " Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Year 1987 - A Report to the Congress of the United States of America," dated February 1986.

2. Letter from David A. Ward, Chairman, Advisory Committee on Reactor Safeguards, to Nunzio J. Palladino, Chairman, U.S. Nuclear Regula-tory Commission,

Subject:

ACRS Comments on the NRC Safety Research Program and Budget for Fiscal Year 1987," dated June 11, 1985.

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l REVIEW AND EVALUATION OF THE NUCLEAR REGULATORY COMMISSION SAFETY RESEARCH PROGRAM FOR FISCAL YEAR 1987 3

5 A REPORT TO THE CONGRESS OF THE UNITED STATES OF AMERICA BY THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS U.S. NUCLEAR REGULATORY COMMISSION l

WASHINGTON, D.C. 20555 FEBRUARY 1986 l

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INTRODUCTION j

In a letter to NRC Chairman Palladino, dated June 11, 1985, the Advisory Committee on Reactor Safeguards (ACRS) commented on a proposed research l

l program for FY 1987 based on program-support funding of $128.6 million.

A copy of that letter is attached. The funding now proposed for FY 1987 is $99 million.

Our coment's herein address chiefly the changes in the research program that have been proposed to accommodate this substantial change in funding level.

The program has been reduced substantially in scope from what we re-viewed and commented on in June 1985. Some studies that should be continued, and will ultimately be required, have had to be stopped or drastically reduced. For the rest, with some exceptions, we believe that most of the important questions or concerns that confront the NRC are being addressed. Our comments for each of the five Decision Units are presented below.

REACTOR ENGINEERING We agree with the proposed funding for mechanical / structural engineering and primary system integrity programs, but are concerned about the decision to terminate the work on the qualification of electrical equipment, as noted below.

4 Electrical Equipment Qualification The current proposal would fund work on aging of equipment the re-liability of which can be assured by periodic maintenance while zeroing out funding aimed at assuring the performance of safety-related elec-trical equipment whose performance in an accident or fire cannot be assured by currently available knowledge. We recommend that the work on electrical equipment and its fire response be continued and funded by deferring some of the work on maintainable equipment.

THERMAL-HYDRAULIC TRANSIENTS Research in this Decision Unit addresses issues that are fundamental to the safe operation of nuclear power plants. Historically, a large portion of the NRC's research resources have been devoted to this area.

While many of the questions of a decade or more ago have been answered, some important issues remain unresolved.

Integral Testing Contrary to the NRC proposal, we believe that the planned follow-on program at the Multiloop Integral System Test (MIST) facility should be completed, even if it is necessary to stretch it out over a longer time 19 1

to accommodate annual funding restrictions. The thermal-hydraulic behavior of the Nuclear Steam Supply System (NSSS) for Babcock and Wilcox (B&W) plants is more complex and not as well understood as that for other pressurized water reactors (PWRs). MIST and associated programs should be continued to raise the level of predictability of the B&W systems to the equivalent of the other PWRs.

The currently proposed research budget would provide funds for the foreign programs, 20/3D and ROSA-IV, at the levels proposed in June 1985. We believe that, to the extent practical, proportional reductions should be made in the funding for these programs.

While plans for a new thermal hydraulic test facility have merit in assuring technical capability into the future, they should not be given precedence over the MIST Program effort. Since a consensus has not been reached on the form of a new facility, it is premature to commit major funding to it at this time. We agree with the proposed funding re-duction.

Separate Effects A viable program in separate-effects testing, especially at universities and smaller laboratories, is in the best, long-term interest of the nuclear safety program. These programs, which can be conducted at a lower cost than any other activity, assure continuing development of the science and scientists necessary for understanding the basic nature of the cooling systems in nuclear power plants. In particular, we believe that visual studies of thermal-hydraulic phenomena, and studies to understand the complicating effects water hammer may have on thermal-hydraulic transients, should be funded.

Models and Codes Development of useful and powerful analytical tools has been a signifi-cant accomplishment of the U.S. nuclear power industry -- especially of the AEC/NRC and their contractors. The present program of code valida-tion, using foreign experimental data, appears to be a cost-effective means to keep these efforts continuing for the benefit of nuclear power plant safety. The development of more user-friendly tools, the nuclear plant analyzer and the data bank, to permit shorter turn-around times in analyzing problems that continue to arise, are useful for both the NRC and industry. Nevertheless, we believe that these latter activities can be assigned a lower priority.

ACCIDENT EVALUATION The current activity in this Decision Unit is .it a point where inte-gration and thoughtful contemplation of previous work is needed to 192

define further investigations. Under these circumstances, we consider the total amount allocated to be acceptable.

Because of the importance of a continuing exploration of the sequences of events that may lead to severe accident consequences, we recomend that the previously prcposed Severe Accident Sequence Analysis program be continued. Necessary funds for this purpose should be reallocated from the source term work.

We do not foresee significant harm to the total program if the work on in-pile fuel behavior is delayed or eliminated.

Although the aerosol work might provide useful information, and might conceivably enable some relaxation of regulatory requirements, currently available information is adequate for regulatory needs.

We are pleased to observe that progress is being made toward a better understanding of containment system performance and toward the develop-ment of associated performance criteria. However, a careful evaluation of the performance of a number of the containment systems that exist in operating plants has still not been made. In view of the importance of the containment system performance as a last and extremely important barrier to the release of radioactive materials in case of a severe accident, we recommend that this work be given a continuing high priority.

REACTOR OPERATIONS AND RISK In our June 11, 1985 report to the Comission, we agreed with the proposed funding level for this Decision Unit but, as we had on several previous occasions, disagreed with the allocation of funding within this Decision Unit. Our reasons are discussed in some detail in our June 11, 1985 letter and in other previous reports on the NRC research program.

The currently proposed funding for this Decision Unit is less than that proposed in June 1985. Although we agree with the NRC proposal to terminate and/or defer programs in the lowest priority areas to accommo-date this reduction, our previous concerns regarding the specific assignment of priorities within this Decision Unit remain.

In our previous reports, we had identified important risk-related licensing problems that are not being addressed in the current and/or proposed research programs. We recomend that part of the funding now proposed for the work on the examination of Technical Specifications be reallocated to the investigation of these problems. The work on the examination of Technical Specifications is important, but is more appro.

priately done by industry. Of the licensing problems identified in previous reports, we recomend that the following three areas be em-phasized:

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• A continued search for possible weaknesses in the current proba-bilistic analyses, e.g., accident paths either not currently evaluated or dismissed as insignificant, which may, 'on closer scrutiny, prove to be very important rto risk,.

• An improved evaluation of the entire family of containmet.t designs, including the effectiveness of possible design improvements.

• The development of improved methods for factoring uncertainty into decision making.

We believe that probabilistic risk analysis (PRA) provides a powerful tool for the rational evaluation of both health and economic risk and its use in the regulatory process should be encouraged. Several utilities are aggressively learning to use this methodology in managing risk and assessing regulatory requirements. The NRC has proposed that funding for the Integrated Safety Assessment Program be. eliminated and that funding for the Office of Nuclear Reactor Regulation's (NRR'is) review of industry PRAs be sharply reduced in FY 1987. We believe that this is a serious mistake and would significantly impact on industry efforts to utilize PRA assessment techniques.

Human Factors We note thtre is no research on human factors in the proposed program.

While we believe such research is needed, we also believe it is appro-priate to await the results of a recently funded study by the National Research Council of the National Academies of Science and Engineering before specific recommendations are made. A program for FY 1988 and future years should then be defined using the results of this study.

WASTE MANAGEMENT, EARTH SCIENCES AND HEALTH Waste Management The Low-Level Radioactive Waste Policy Amendments Act of 1985 requires that radioactive wastes containing radionuclide concentrations above Class C must be placed in a disposal site supervised by the Department of Energy. The NRC has been assigned the responsibility for licensing such facilities. This Act also requires that within 12 months the NRC must identify disposal methods other than shallow-land burial, and that within 24 months the NRC must provide technical information for states to proceed with alternate disposal practices. We believe that the increased funding proposed is the minimum needed to permit the NRC to perform the research necessary to meet these requirements.

For the program on high-level wastes, we endorse the increased attention being directed to tuff and salt as potential repository media. However, we also recommend that the NRC Research Staff develop a more rigorous 194

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approach and/or, methods for determining the priorities assigned to individual research projects within the high-level waste area. To enable the NRC to conduct adequate research on these media and to meet the responsibilities specified in the Nuclear Waste Policy Act, we again encourage the Congress to consider providing the NRC with additional resources through t,he Nuclear Waste Management Fund.

Earth Sciences In its original budget proposal, the NRC provided funding to transfer to the United States Geological Survey (USGS) the responsibility for operating that portion of the seismographic network in the Eastern United States now supported by the NRC. The agreement with the USGS provided for payment of $5 million over a 5-year period, beginning with

$1 million in FY 1_987 The NRC now proposes that this funding transfer be eliminated to conserve research funds. The NRC also proposes that all other NRC support for the operation of this network .be discontinued after FY 1987. -We believe that continued operation of the seismographic network would be useful, but agree that the NRC has higher priority research for its limited funds. It is our hope that the USGS will be able to continue the operation of this network without NRC support.

l Radiation Protection and Health Effects The NRC has proposed the elimination of funding to the National Academy of Science Committee on the Biological Effects of Ionizing Radiation, to the National Council on Radiation Protection and Measurements, and to the International Comission on Radiological Protection. Since these are the premier organizations that interpret and evaluate data on the health: effects of radiation, we believe that such action is unwise. In our opinion, it is essential that the NRC maintain liaison with and keep abreast of the groups both in the U.S. and abroad who are active in the

fields of radiation protection and health effects. We strongly urge that the NRC continue to provide funds to these organizations.

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UNITED STATES NUCLEAR REGULATORY COMMISSION l

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. wAsemnow. o. c. noses June 11, 1985 l

l l Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Comission Washington, D.C. 20555 i

SUBJECT:

ACRS COMENTS ON THE NRC SAFETY RESEARCH PROGRAM AND BUDGET FOR FISCAL YEAR 1987 <

During its 302nd meeting, June 6-8, 1985, the Advisory Comittee on Reactor Safeguards completed its review of the proposed program i

and associated budget for the Office of Nuclear Regulatory Research (RES) for Fiscal Year 1987. This matter was considered also by the ACRS Subcommittee on the Safety Research Program at meetings on May 8 and June 5, 1985 and at several meetings of other ACRS Subcommittees having interest in specific portions of the research program. During our review, we had the benefit of discussions with representatives of RES, the Office of Nuclear Reactor Regulation (NRR), and the Office of Nuclear Material Safety and Safeguards (255). We also had the benefit of the documents referenced.

Our consnents and recommendations are directed to the budget presented to us by RES on June 5, 1985. On that date, the proposed budget for program support was $128.6 million, distributed among the five Decision Units as shown in the attached table. We note that this figure is $7.6 million (about 6 percent) greater than the $121 million requested of Congress for FY 1986. Although we have some concern about proposing a research program at a level of funding that may be difficult to obtain, we have not tried to agree on recommendations that would reduce the program to a level consistent with the requested budget for FY 1986.

Our comments are provided below and our budget recomendations are sumarized in the attached table.

REACTOR ENGINEERING We endorse the programs and the proposed funding level of $44.6 million for this Decision Unit. If absolutely necessary, some reductions could be made by stretching out programs in some areas.

However, we do not believe that this should be done for the research programs related to " plant aging"; any reductions in these programs would delay the timely completion of this important work or reduce its scope.

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Honorable Nunzio J. Palladino June 11, 1985 THERMAL-HYDRAULIC TRANSIENTS We support the proposed total budget of $23.4 million for this

. Decision Unit.

We believe that a major capability for thermal-hydraulic research should be maintained. It is .especially important to ensure the i

l continued availability of experienced specialists. We support the concept for a new or modified integral test facility to be co-located with a technical staff capable of analysis and code development. The budget proposal of $4.5 million for this effort is appropriate if firm planning for such a facility is begun immediately and if it leads to a decision to proceed.

A vigorous experimental program in Separate Effects should be sponsored and maintained by the NRC at universities and other smaller

] facilities throughout the country and a funding level of about $4 million would be appropriate rather than the $3.2 million proposed by RES. This program should include investigation of phenomena, such as water hammer, which could impact nuclear safety but have not yet been extensively studied in that context. We recommend that a program of ,

flow visualization studies, proposed earlier by RES, be reinstated.

The advanced thermal-hydraulic codes, such as TRAC and RELAP, have contributed greatly to an understanding of reactor operation and

risk. To ensure their continued availability and utility, they must be effectively maintained. We endorse the NRC program to maintain these codes and to continue their assessment and improvement by a program of cooperation with foreign research institutions. We caution, however, that a strong domestic capability for maintaining and using these codes must continue.

The nuclear plant databank (NPDB) and nuclear plant analyzer (NPA) programs are intended to extend the utility of the major thermal-hydraulic codes. The NPDB program, while important to users in both NRR and RES, is not research. We recommend that it be funded from the NRR budget. The NPA program has the potential for providing an extremely flexible and useful tool for nuclear systems analysts.

We believe this development program is important and should be con-tinued. However, one part uf the NPA program involves development of parallel-processing using snialler, special-purpose computers as a i substitute for mainframe computers used for the major

. thermal-hydraulic codes. While this work is interesting and may be of general use, it is not directly related to the NRC's thermal-hydraulic research program. We recommend that it be funded elsewhere.

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l Honorable Nunzio J. Palladino June 11, 1985 Considering our recomendations above, we believe that funding for

> the Transient Models and Codes program can be reduced from the l proposed $5.8 million to about $5 million.

ACCIDENT EVALUATION I With publication of NUREG-0956, the source term reassessment report, the NRC Staff will have completed the first phase of the development of a suite of codes meant to describe the course of a severe accident with sufficient detail and accuracy to permit the results of the calculational methods developed to be used in regulatory decision making. In recent meetings with the RES Staff, we have been assured that, in their view, the results of the calculational methods described in NUREG-0956 can, indeed, be used for that purpose.

Nevertheless, RES proposes further research aimed at refining the methods and removing uncertainties in the approach described in the source term reassessment report. This is said to be partly in response to coments contained in the recently released review by the American Physical Society of the source term research results.

However, the deficiencies said to exist, and the uncertainties that will be removed by the proposed research for FY 1987 have been described only qualitatively. Before more research can be justified, deficiencies in existing codes should be identified, a more detailed and quantitative description of what is to be done to remove these deficiencies should be developed, and the improvements expected and the effect they will have on the regulatory process should be described.

Although we believe that the course of action recomended will improve the likelihood of obtaining useful results, we note that we have repeatedly expressed skepticism concerning placing almost total dependence on the results from large mechanistic codes as a tool for decision making. We continue to recomend that any decisions made take due account of the large uncertainties in the results of so-callad mechanistic codes.

In NUREG-1105, we recomended initiation of a program to obtain information needed to develop containment perfonnance criteria.

That has not occurred. Nor does support for such a program appear in the proposed FY 1987 budget. *We urge that a program be initiated with funds coming from the Fission Product Source Term work.

We have, within the past few days, received the coments of NRR on the proposed severe accident research program. We share NRR's doubts concerning the wisdom of continuing work on MELPROG. We are, however, pleased to observe that there appears, finally, to be an effort to evaluate the use to be made of results from the research program related to severe accidents, and to design future research to provide whatever additional information may be needed by NRR.

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Honorable Nunzio J. Palladino June 11, 1985 Although we are not as sanguine as NRR appears to be about the adequacy of the infonnation that will be available by the end of FY 1986, we believe that the process of negotiation which seems to have begun between NRR and RES should define a more useful research

program than we have seen described to date. We recomend that $2.5 million of the proposed funding be reallocated from those portion; of the research dealing with in-pile tests to support research on human factors and man-machine interaction. We cannot coment further on the' proposed budget for this Decision Unit until we have the results '

of the negotiations between NRR and RES.

REACTOR OPERATIONS AND RISK In NUREG-1105, we took strong exception to the NRC decision to i terminate research in the human factors area in FY 1986. We deeply regret that RES persists in omitting such research in its proposed FY 1987 budget. We strongly recommend that this decision be recon-sidered and that a carefully formulated long-range research program in human factors and man-machine interaction be launched. We recom-mend that about $2.5 million be taken from the in-pile experimental program on Damaged Fuel and Source Tenn and reallocated to research on human factors and man-machine interaction. This should include matters such as design, maintenance, quality of personnel, and basic management philosophy, as well as coping with unexpected complex transients.

We agree with the proposed total level of support of $15.6 million for risk-related programs in this Decision Unit, but do not agree with the allocation to various tasks or with all of the tasks themselves.

In NUREG-1105, we recommended that a thoughtful research effort be devoted to a search for possible weaknesses in current probabilistic analyses, e.g., accident paths either not currently evaluated or dis-

missed as insignificant, which may, on closer scrutiny, prove to be very important to risk. We reiterate that recommendation here. An example of a subject for such examination might be whether pressure components whose gross failure probability is assumed to be negligi-ble are designed, fabricated, operated, and inspected with validated techniques to an extent commensurate with the low assumed failure u probability. We believe that such research is properly the role of  ;

' the NRC and not to undertake it reflects an ostrich-like position.

We also recommend the initiation of, or increased emphasis on, ,

research on methodology for treating design errors, systems interactions, plant aging, equipment qualification, and external

floods in PRAs. ,

l l

200

Honorable Nunzio J. Palladino June 11, 1985 Furthemore, we believe that in FY 1987 the research on evaluation of severe accident risk and risk management is being curtailed or even terminated prematurely. Issuance of the proposed policy statement on severe accidents will not tell us how to specify a containment performance criterion. It will not tell us the importance of plant-specific features to risk. Nor will it tell us the role of external events, design errors, human cognitive errors, etc., in evaluating risk. We recomend continued strong support for studies having the appropriate focus.

We believe that the proposed program on regulatory and inspection applications as well as the plant-specific calculations with MELCOR should De reduced or stretched out as necessary to pemit adequate support of the risk-related programs recomended above.

WASTE MANAGEMENT, EARTH SCIENCES, AND HEALTH We believe that the proposed funding for the programs on High-Level Waste Management and Low-Level Waste Management is adequate. A similar conclusion has been reached regarding the funding for research on Health Effects, exclusive of the needs for support on occupational radiation protection.

As before, we recomend that a research program aimed at improving our knowledge of the likelihood of severe floods should be undertaken. Support for such research (about $0.25 million) should be obtained by reallocation of funds within this Decision Unit.

RES has proposed that all support for work on occupational radiation protection be eliminated from the FY 1987 budget. We do not agree.

Examples of problems that need to be addressed include the development of:

a. A radiation protection program data base;

b. A standard methodology for optimization analyses;

c. A system of performance indicators for assessing the qual-ity of nuclear power plant radiation protection programs; and

d. Improved clothing for protection against internal exposures, including studies of the net risks of using respirators.

201

Honorable Nunzio J. Palladino June 11, 1985 It is estimated that items "a" and "d", which have been endorsed by NRR, would require $0.5 million for FY 1987. Support of items "b" and "c" is estimated to require 30.2 million. We urge that funds for the support of these activities be provided either within the NRC research budget or through Technical Assistance Programs within NRR.

Sincerely, David A. Ward Chainnan

Attachment:

Table 1 - Office of Nuclear Regulatory Research Program Support Budget for FY 1987.

References:

1. Table entitled, " Nuclear Regulatory Research FY 1987 Budget,"

dated June 4, 1985, submitted at the June 5, 1985 ACRS Safety Research Program Subcomittee meeting

2. Memorandum from Robert B. Minogue, Director, Office of Nuclear Regulatory Research, to Jack W. Roe, Chairman, Budget Review Group,

Subject:

NRR's Comments on the RES Budget, dated May 31, 1985

3. Memorandum from Harold R. Denton, Director, Office of Nuclear Reactor Regulation, to Robert B. Minogue, Director, Office of Nuclear Regulatory Research,

Subject:

NRR Review of RES FY 1987 Internal Budget Presentation to ACRS and BRG, dated May 29, 1985

4. Memorandum from E. F. Conti, Acting Director, Policy, Planning and Control Staff, Office of Nuclear Regulatory Research, to L.

W. Barry, Director, Office of Resource Management.

Subject:

Analysis of "Hard Call" Programs in the RES FY 1987 Budget Allocation, dated May 21, 1985

5. Memorandum from E. F. Conti, Acting Director, Policy, Planning and Control Staff, Office of Nuclear Regulatory Research, to RES Division Directors,

Subject:

Considerations Related to the FY 1987 Allocation to the $13D Million EDO Mark, dated May 17, 1985

6. Memorandum from E. F. Conti, Acting Director, Policy, Planning and Control Staff, Office of Nuclear Regulatory Research, to L.

W. Barry, Director, Office of Resource Management

Subject:

FY 1987 Internal Review, dated May 17, 1985

7. Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Comission, " Reassessment of the Technical Bases for Estimating Source Terms," Draft NUREG-0956, dated April 1985 202

l June 11. 1985 TABLE 1 0FFICE OF NUCLEAR REGULATORY RESEARCH PROGRAM SUPPORT 8UDGET FOR FY 1987 (DOLLARSINMILLIONS)

RES REQUEST ACRS DECISION UNITS (JUNE 5, 1985) RECOMMENDATIONS

1. REACTOR ENGINEERING $ 44.6 $ 44.6

2. THERMAL-HYDRAULIC TRANSIENTS 23.4 23.4

3. ACCIDENT EVALUATION 27.4 24.9

4. REACTOR OPERATIONS AND RISK 15.6 18.1

5. WASTE MANAGEMENT, EARTH SCIENCES, AND HEALTH 17.6 17.6 TOTAL $128.6 $128.6 203

p2 H4

% 9'o,, UNITED STATES T n NUCLEAR REGULATORY COMMISSION

$ .E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS WASHINGTON, D. C. 20555 O,

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%,*****s ,4 June 11, 1986 Honorable Nunzio J. Palladino Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555

SUBJECT:

ACRS COMMENTS ON THE NRC SAFETY RESEARCH PROGRAM AND BUDGET FOR FISCAL YEAR 1988 During its 314th meeting, June 5-7, 1986, the Advisory Committee on Reactor Safeguards completed its review of the proposed program and associated budget for the Office of Nuclear Regulatory Research (RES) for Fiscal Year 1988. This matter was considered also .by the ACRS Subcommittee on the Safety Research Program at meetings on May 7 and June 4, 1986 and at several meetings of other ACRS subcommittees having interests in specific portions of the research program. During our review, we had the benefit of discussions with representatives of RES, the Office of Nuclear Reactor Regulation (NRR), and the Office of Nuclear Material Safety and Safeguards (NMSS). We also had the benefit of the documents referenced.

Our coments and recommendations are directed to the program presented to our Subcommittee on June 4,1986, based on an allocation of $99.6 million for program support. Although the program presented to us was categorized in accordance with Mission Area Codes (MACS), our comments are directed to the traditional program areas.

Our comments are provided below. In general, they relate to research that is not being proposed but should be, or in some cases to research that is proposed but may not be needed. For those programs that are not mentioned, we are in general agreement with what the Staff is proposing.

REACTOR ENGINEERING Primary System Integrity Cast stainless steel components in the primary system lose ductility with time in service. The implications of this phenomena for long-term 205

Honorable Nunzio J. Palladino June 11, 1986 primary system integrity need more attention. Appropriate emphasis should be placed on ascertaining the likelihood of flaws resulting from fabrication or from service, and on developing means of assessing conditions under which they could pose significant risk. i Electrical Equipment Qualification The research and funding for Electrical Equipment Qualification proposed for FY 1988 are inadequate. In our report dated February 19, 1986 to the Congress, we recommended that this research be continued; contrary to this recommendation. the research is scheduled for termination by the end of FY 1986 and funding has been eliminated for FY 1987 and there-after. We consider the continuation of this work to be vital to the nuclear safety program and again recommend that it be funded.

The objective of this research is to assess the probability of survival and performance 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 vital to preventing accidents as well as mitigating the consequences of acci-dents, should they occur.

Four unique test facilities, with a combined cost of over $2 million, were constructed at Sandia especially for this program. These facil-ities are: (a) the high-intensity adjustable cobalt radiation facility for creating an accident environment of steam, radiation, chemicals, and gas; (b) the low-intensity adjustable cobalt facility for creatin temperature-radiation environment for life aging of specimens; (c) gthe a high-temperature facility for creating severe accident environments; and, (d) the fire test research facility for assessing the effects of fires, including hydrogen burns. These test facilities are unique in being able to create environments for assessment of the effects on aged equipment of practically any credible incident, including design basis or severe accidents, hydrogen burns, and fires. In order to preserve and maintain the existing test facilities and staff experience suffi-cient to continue and complete the electrical equipment qualification work efficiently, it will be necessary to continue funding this work in FY 1987. Since its inception from about 1976, more than $10 million has been spent on this program. Funds needed to continue and complete the program are approximately $1.5 million in FY 1987 and $0.9 million in FY 1988.

Electrical equipment qualification work and the electrical elements in the program to determine the effects of plant aging and irradiation are 206 l

Honorable Nunzio J. Palladino June 11, 1986 closely related and to some extent interdependent. The research on aging in this program can utilize the same testing facility and exten-sive experience base for the electrical equipment qualification work.

Both programs are essential as input to the program to determine life expectancy of electrical equipment for future reactor life extension licensing.

We recommend that the Electrical Equipment Qualification work that was originally included in the FY 1987 budget be completed and be funded by deferring or stretching out other work, for example, the program to determine the effects of aging and irradiation.

THERMAL HYDRAULICS Research planned in the thermal-hydraulics area is divided into two general parts. The first is a continued program to improve the under-standing 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 cxperiments which does not include direct industry participation.

We believe that the emphasis on B&W systems is appropriate and that the industry should provide major support to this program as the RES Staff has proposed. It is also important that the NRC continue a viable general program of research into the foreseeable future, including integral testing, separate-effects testing, and code development.

The following comments relate to research areas for which neither the program plans nor the funding levels have been established clearly.

Multi-Loop Integral System Test (MIST) Follow-On Substantial questions remain about the thermal-hydraulic behavior of B&W systems. There is no independent program by the vendor or utilities to address these questions. We believe that a limited series of tests should be conducted following the initial program. However, it is not clear to us that the " full power" tests are justified. Additional funding, including substantial support from the industry, is essential.

Once-Through Steam Generator (OTSG) Separate Effects Test The OTSG is a unique feature of B&W nuclear steam supply systems. The transient behavior of the OTSG, especially with auxiliary feedwater flow, cannot now be predicted with accuracy. A large-scale experimental model of an OTSG proposed by RES can provide needed information. This facility would complement the MIST follow-on program noted above. It 207

Honorable Nunzio J. Palladino June 11, 1986 is not necessary that this facility be capable of full pressure, but it should be large enough in scale to permit accurate modeling and measure-ment of key phenomena.

Water Hammer We believe that it is essential to continue development of an experi-mental and analytical program to provide a better understanding of the causes and consequences of water hammer in nuclear power plants, espe-cially as a complicating effect in complex transients induced from other Causes.

Bleed and Feed Phenomena The ability of pressurized water reactor (PWR) plants to " bleed and feed" from the primary system, either to remove core heat or to permit controlled depressurization of the primary coolant, is assuming in-creased importance as more attention is being paid to procedures and methods for managing off-normal and accident situations. In addition, increased consideration is being given to the design of systems specifi-cally for controlled primary blowdown. Largely because of the locations of valves and pumps used for this process, the thermal-hydraulic phenomena involved in bleed and feed can be quite complex. The phenome-na involved in the process and the capability of equipment called on are not well enough understood to validate the designs and to support the development of applicable rules, procedures, or practices. We believe that a research program, possibly in the proposed new integral test facility, should be developed to provide an appropriate base of informa-tion. Separate-effects tests and equipment tests may also be necessary.

SEVERE ACCIDENTS The research program being proposed under the rubric of Severe Accidents includes a significent experimental component. The relationship of this research to dealing with the severe accident regulatory issue should be made clearer than it now is. For example, there are three major experi-mental 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 interaction, and to containment behavior under extreme over-pressure. Presumably each of these experiments is designed to reduce the uncertainties estimated in connection with the preparation of NUREG-1150, " Nuclear Power Plant Risk and Regulatory Applications." In order to decide whether additional research is needed to reduce existing uncertainties, NRR must decide what uncertainties are acceptable in connection with its regulatory responsibilities. We have not seen any 208

Honorable Nunzio J. Palladino June 11, 1986 indication that this decision has been made. We (and we would suppose RES as well) must, under the circumstances, try to judge the relevance of the proposed research with insufficient information.

With this caveat, we make the following comments:

• The research on containments under extreme overpressure seems well designed and should produce results that are relevant and that are needed to calibrate codes being developed for a description of containment behavior.

• The proposed work on core-concrete interaction is needed. However, in order to plan research that will answer questions likely to arise in regulation, the important uncertainties attributable to incomplete understanding of particular phenomena should first be identified. The research can then focus on those phenomena.

• Some risk analyses indicate that high pressure core melt sequences may produce enough direct heating of containment atmosphere to cause early containment failure by overpressure. The proposed research on direct containment heating may be of considerable significance to understanding risk; however, it may be relevant only if a high pressure core melt sequence for PWRs has a suffi-ciently high probability. We believe that the proposed research be performed concurrently with an investigation of the likelihood and effect of direct-heating events. We also recommend that the experimental investigation give first consideration to the expul-sion process, to the effects of containment geometry, and to the presence of water in the sub-vessel cavity.

These comments do not represent a complete coverage of the severe accident research program. We use them as an example of our conclusion that more consideration 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 another way, what are the questions that regulators have encountered or are likely to encounter in formulat-ing regulations for dealing with severe accidents that cannot be answered with existing information. With diminishing resources, it is increasingly important that the research be specifically designed to address safety concerns.

RISK ANALYSIS We agree with the proposal for increased emphasis on work related to understanding the risk significance of equipment qualification, aging and fire. We recommend that a continued search for possible weaknesses 209

Honorable Nunzio J. Palladino June 11, 1986 in the current 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.

(

It is important that a dependable assessment be made of the containment performance for each containment type for a wide range of postulated scenarios. We recommend that research on the development and evaluation of containment performance criteria be continued. We also recommend that research into the management of severe core damage accidents and on containment performance be given emphasis. It is vital to evaluate each type of nuclear steam supply system and the major variations in contain-ment design. Some research will also be required on complex systems interactions if this area is to be appropriately treated in the NRC's severe accident evaluation.

HUMAN FACTORS We anticipate that the Na tional Research Council will recommend a

• considerable program of research effort for human factors. Plans for implementation of such a program should be factored into the budget.

WASTE MANAGEMENT High-Level Waste We support the efforts of the NRC Staff to develop an improved system for assigning priorities to individual research projects pertaining to High-Level Waste Management. We believe that more emphasis should be directed to the development of field data on the movement of radio-nuclides within the environment and on the associated impact of heat-water-rock interactions. There is also a need for the development of data on the predicted performance of repository systems under realistic field conditions. Although to a major extent such work should be conducted by the Department of Energy (DOE), the NRC Staff needs to be sufficiently active to provide the Commission with a capability for independent assessment of the DOE proposals.

Low-Level Waste We endorse the program on Low-level Radioactive Waste Management. At the present time, the NRC is one of the few Federal agencies conducting research in this field. The results of these studies are urgently needed by the states which have been mandated by the Congress to meet a strict wastes.

timetab,le for developing facilities for the disposal of such We believe that the program, as outlined, is the minimum necessary to meet NRC responsibilities in this area.

210

June 11, 1986 I Honorable Nunzio J. Palladino Additional funds should be provided for research to develop a technical basis for defining criteria for the designation of low-level radioactive materials that are below regulatory concern, and for radiation protec-tion guidance for alternatives to shallow-land burial of low-level wastes. Both of these needs are mandated by the Low-Level Radioactive Waste Policy Act of 1985.

Wastes with radionuclide concentrations above Class C are not receiving adequate NRC research attention. Since the Low-Level Radioactive Waste Policy Act of 1985 requires that DOE develop methods for the management of such wastes, and since the related disposal facilities must be licensed by the NRC, we urge that the NRC move promptly to define and initiate a research program to develop an appropriate technical base for meeting its needs in this subject area.

Sincerely, 80, du 4 David A. Ward Chairman

References:

1. Table on the Office of Nuclear Regulatory Research Program Support Budget for FY 1986-1989, dated June 3, 1986, submitted at the June 4, 1986 ACRS Safety Research Program Subcommittee meeting.

2. U.S. Nuclear Regulatory Commission, Advisory Committee on Reactor Safeguards, " Review and Evaluation of the Nuclear Regulatory Commission Safety Research Program for Fiscal Year 1987 - A Report to the Congress of the United States of America," dated February 19, 1986.

3. U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, " Reassessment of the Technical Bases for Estimating Source Terms," Draft NUREG-0956, dated May 23, 1986.

211

o UNITED STATES

-~ NUCLEAR REGULATORY COMMISSION j ;E ADVISORY COMMITTEE ON REACTOR SAFEGUARDS g W ASHINGTON, D. C. 20555 o,

1. '+9 go May 13, 1986 Honorable Nunzio J. Palladino Chairman U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Dear Dr. Palladino

SUBJECT:

ACr.5 COMMENTS ON THE DEFINITION OF LOW-LEVEL RADI0 ACTIVE WASTE According to the Low-Level Radioactive Waste Policy Amendments Act of 1985, the Nuclear Regulatory Comission is authorized to classify certain materials as low-level radioactive waste.

This being the case, we believe this law provides the NRC an excellent opportunity to bring within the low-level radioactive waste definition a.

category of wastes which heretofore has not been properly addressed or controlled. These are the so-called " Naturally Occurring and Accel-erator Produced Radioactive Materials," comonly referred to as NARM wastes.

We recomend that you encourage the NRC Staff and the Commission to expand the definition of low-level radioactive waste to include those categories of NARM wastes that are important to public health and safety. Such a change would bring under regulatory control a category of radioactive waste that for too long has been neglected.

Sincerely,

. b.

David A. Ward Chairman 213

, o UNITED STATES g c, NUCLEAR REGULATORY COMMISSION ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

!")M,/(//

S

• k j

~E WASHINGTON, D. C. 20555 g * *v

• • • August 13, 1986 l

Mr. Victor Stello, Jr.

Executive Director for Operations ,

U.S. Nuclear Regulatory Commission Washington, D. C. 20555

SUBJECT:

ACRS COMMENTS ON VARIOUS NMSS AND RES WASTE MANAGEMENT TOPICS During its 316th meeting, August 7-9, 1986, the Advisory Comittee on Reactor Safeguards heard a report from its Subcomittee on Waste Management regarding the topics listed below. Each of these topics was reviewed by the Subconmittee during a meeting on July 21-23, 1986, at which presentations were made by representatives of the Nuclear Regulatory Commission and the Environmental Protection Agency.

1. NMSS radioactive waste management program, including the Division of Waste Management's five-year plan, the proposed Federally Funded Research and Development Center (FFRDC), the proposed use of rulemaking to bring key prelicensing issues to closure, and alternative methods to shallow land burial.

2. Programs of the Waste Management Branch, Office of Nuclear Regulatory Research (RES), including the development of field data on the movement of radionuclides within the environment and the associated impact of heat-water-rock interactions, and the predicted performance of repository systems under realistic field conditions.

3. Generic technical positions on " Determination of Radionuclide Sorption" and " Determination of Radionuclide Solubility" for high-level nuclear waste repositories.

4. Development of residual radiation limits for the disposition of

' land, buildings, equipment and metals resulting from the decontamination and decomissioning of nuclear power plants and fuel facilities.

5. Salvaging of contaminated smelted alloys containing technetium-99 and/or low-enriched uranium as residual radioactive contamination.

6. NRC Staff policy statement and implementation of NRC policy on radioactive wastes "below regulatory concern."

Subcomittee reports on each of these topics are attached. Key recomendations included in these several reports and endorsed by the full Comittee are:

215

1 l

Mr. Victor Stello, Jr. 2 August 13, 1986

1. Although we endorse the Staff's proposed use of rulemaking as a means to bring key prelicensing issues to closure, we recommend that the NRC Staff develop a statement outlining the logic by which this approach is being formulated, why it is considered viable, details on how it is to be implemented, and the time required for its implementation.

2. We believe that the establishment of the proposed FFRDC will be helpful. However, the progress of work at the Center should be carefully monitored to assure that it is accomplishing its intended goals.

3. We support the establishment of an adequate system of peer review to assure the credibility of the waste management activities of the RES and NMSS staffs. It is important, however, that such reviews be conducted by groups that include people who are experienced and knowledgeable, and who will be able to provide comments reflecting

, a spectrum of views and technical positions.

4. The Generic Technical Positions being developed on radioactive waste management should emphasize performance goals rather than being prescriptive.

5. We continue to believe that separate criteria and standards must be developed for the release of land and fixed facilities, and for the release of equipment and materials, for general public use.

However, this effort should proceed in a coordinated manner.

6. The release of contaminated materials from enrichment plants should be considered a part of the larger generic question, including items such as de minimis concentrations, wastes "below re concern" (item 7, below), and residual radiation limits (gulatory item 5, above).

7. We endorse the cooperation of the NRC Staff with the EPA on the development of standards for radioactive wastes "below regulatory concern" and related matters.

The ACRS discussed and approved transmittal of the subcomittee reports for your consideration.

Sincerely,

, k, David A. Ward Chairman Attachments:

As stated 216

Mr. Victor Stello, Jr. 3 August 13, 1986

References:

Provided in attached reports cc: Chairman Zech Commissioner Roberts Commissioner Asselstine Commissioner Bernthal D. F. Ross, Acting D/RES K. R. Goller, RES/DRA R. Alexander, RES/SRAB G. Arlotto, RES/D/ DES F. Costanzi, RES/WMBR J. Davis, D/NMSS D. Mausshardt, DD/NMSS R. Browning, NMSS/DWM M. Bell, NMSS/DWM 217

August 8, 1986 Report No. 1 ACRS Waste Management Subcommittee Comments on NMSS Radioactive Waste Management Program July 21-23, 1986

1. The Subcommittee reviewed the plans of the NMSS Staff for implementation of the Five-Year Plan for the High-Level Waste Program (Reference 1).

The Subcommittee was encouraged by, and fully endorses, the active tenor of the program. We believe this is a desirable step forward, and we urge that the NRC Staff assure that they have the necessary technical depth to support the program. Establishment of the Federally Funded Research and Development Center (Reference 2) may assist in this regard.

With respect to this Center, however, the Subcomittee offers the following suggestions,

a. The progress of the Center should be carefully monitored (as we understand it will be), especially during the early years, to assure that it is accomplishing its intended goals.

b. A significant part of this monitoring effort should be to assure that the Center is able to attract and retain experts who are recognized as being competent in the technical fields pertinent to radioactive waste management. This goal will require stabilization of funding to the extent possible.

2. Although we endorse the Staff's use of rulemaking as a means for bring-ing key prelicensing issues to closure (Reference 3), considerable care must be exercised in the selection of such issues to assure that they pertain to fundamental principles that must be established in order to move forward with the licensing of a high-level waste repository. We concur that the method for assuring that the performance of a repository will meet the EPA standards is an excellent example of such an issue.

However, the Subcomittee would like to be provided with the logic by which this approach is being formulated, why it is considered viable, details on how it is to be implemented, and the time required for its implementation. This information should be prepared in a format on which the Subcomnittee and Comittee can comment.

3. In terms of the need to develop regulations for alternative methods to shallow land burial for the disposal of low-level wastes (Reference 4),

we recomend that Oe NRC Staff:

a. Rall the States to determine which alternatives they prefer. The responses should be helpful in reducing the number of approaches that need to be evaluated.

218

Radioactive Waste Management 2

b. Meet with EPA Staff members to solicit their suggestions and recommendations for selecting which alternatives to consider.

Once this information has been assembled, we suggest that the NRC Staff group the alternatives so that applicable disposal criteria can be developed on a generic basis.

References:

1. Division of Waste Management High-Level Waste Program Five-Year Plan, FY86-FY90, undated.

2. SECY-86-192, Policy Issue (Notation Vote), Sponsorship of a Federally Funded Research and Development Center (FFRDC) for Waste Management Technical Assistance and Research (SECY-85-388), dated June 27, 1986.

3. Presentation Handout, Early Identification and Closure of Licensing Open Items, J. Linehan, Division of Waste Management, NMSS, dated July 22, 1986.

4. Presentation Handout, Summary of NRC's Work on Alternative Disposal Methods to Shallow Land Burial, M. Knapp and C. Pittiglio, dated July 21, 1986.

219

August 8, 1985 Report No. 2 ACRS Waste Management Subcommittee Comments on Programs of the Waste Management Branch Office of Nuclear Regulatory Research July 21-23, 1986

1. The Subcommittee was impressed with the review being conducted of natural systems that might serve as analogs for various processes that are anticipated to occur within a high-level radioactive waste repository. We endorse these efforts.

2. The Subcommittee concurs that an adequate system of peer review needs to be established to assure the credibility of the waste management activ-ities of the RES Staff. We encourage the RES Staff to continue to explore the development of such a system. Possibilities that should be explored include:

a. Developing Staff positions on certain issues through the prepara-tion of appropriate Regulatory Guides or Branch Technical Po-sitions.

b. Permitting outside comment on certain issues through the presenta-tion of proposed RES Staff positions through the mechanisms of workshops and similar forms of public technical meetings.

c. Considering the establishment of peer review groups that can be rapidly convened and can provide prompt comments, as needed, on certain key issues. In the opinion of the Subcommittee, the use of professional societies and trade organizations to perform such functions, although sound, could encounter such delays as to seriously hamper the usefulness of the resulting comments.

d. Exercising care to assure that peer review groups include people who are experienced and knowledgeable, and who will be able to provide comments reflecting a spectrum of views and technical positions.

References:

1. Presentation Handout, Repository Performance Prediction Under "Realis-tic" Field Conditions, T. J. McCartin, Waste Management Branch, RES, dated July 23, 1986

2. Presentation Handout, Radionuclide Movement and Heat-Water-Rock Inter-actions in the Natural Environment, L. A. Kovach, Waste Management Branch, RES, dated July 23, 1986 l

220

Nuclear Regulatory Research 2

3. Presentation Handout, Long-Term Performance Demonstration, F. A.

Costanzi, Waste Management Branch, RES, dated July 23, 1985 I

df

August 8, 1986 Report No. 3 ACRS Waste Management Subcommittee Comments on Generic Technical Positions on " Sorption" and " Solubility" July 21-23, 1986

1. On the basis of its review of the written reports and discussions with the NRC Staff, the Subcommittee believes that the two documents refer-enced below are too prescriptive and that they cover topics that are not germane to the regulatory role of the NRC. An example of the former is the specification of the matrix of experiments to be developed as a planning tool for characterizing the sorption properties of a subsurface repository. An example of the latter is the concern expressed that DOE may not conduct related experiments in the most expeditious manner.

2. Using the two documents cited as a basis, we recommend that the NRC Staff move forward to develop documents that will be more suitable to its needs as well as to those of the DOE. In preparing these reports, the NRC Staff should direct its primary attention to the specification of the " products" required from the DOE to support their licensing application, not to the manner or mechanisms through which these

" products" are to be obtained. As is clearly stated in the opening paragraph of Reference 1, primary attention should be directed to the

" approach" for determing such solubilities, not to the prescription of

" methods" for making such determinations.

References:

1. Determination of Radionuclide Solubility in Groundwater for Assessment of High-Level Waste Isolation, Technical Position, Geotechnical Branch, Division of Waste Management, dated November, 1984

2. Determination of Radionuclide Sorption for High-level Nuclear Waste Repositories, Draft Technical Position, Geotechnical Branch, Division of Waste Management, dated January, 1986 222

August 8,' 1986 Report No. 4 ACRS Waste Management Subcommittee Comments on Development of Residual Radiation Limits for the Disposition of Land, Buildings, Equipment and Metals Resulting from the Decontamination and Decommissioning of Nuclear Power Plants and Fuel Facilities {

July 21-23, 1986

1. The Subcommittee is pleased to note the cooperation being exercised by the NRC and EPA Staffs in the development of guidance and standards related to this subject (Reference 1).

2. We continue to believe (Reference 2) that:

a. It would be best to separate the development of criteria and standards for the release of land and fixed facilities from those developed for the release of equipment and materials for general public use. Because of the complicated nature of and certain differences in the issues involved, such separation -

may prove necessary. L

b. At the same time, however, we believe that these two problem areas need to be addressed in a coordinated manner. In all probability, the decommissioning and decontamination of fixed facilities will result in many items suitable for consid-eration for release to the public, prior to the fixed facil-ities being made ready for general public access.

3. The Subcommittee suggests that data be assembled and analyzed on the criteria and standards used in the past in resolving similar questions pertaining to related facilities and sources. Examples include earlier guidance by the Federal Radiation Council, cleanup standards for inactive uranium mill tailings sites, protection guidance developed for phosphate lands, and similar guidance being applied in the Formerly Utilized Sites Remedial Action Program (FUSRAP). We understand EPA is preparing such an analysis; we endorse this effort.

4. This subject is closely related to other topics now under study, e.g., levels of radioactivity "below regulatory concern" and the disposal of scrap slightly contaminated with technetium-99 and low-enriched uranium. We urge that the NRC Staff develop generic cniteria that would be broadly applicable in defining the risk that would be acceptable to a large population from such activities and therefrom to provide estimates of the corresponding levels of residual radioactive material or contamination that would be acceptable.

5. Certain factors need to be considered for the potential exposure scenarios and models that are being developed to estimate the 223

RADIATION LIMITS... 2 population doses through each pathway. For each such model there is a need to specify the uncertainties that are acceptable to determine the realism or conservatism in the resulting dose esti-mates and to agree on an acceptable procedure for the necessary validations. The goal should be to provide as realistic an assess-ment as is practical. We understand such models are being devel-oped at the Pacific Northwest Laboratories (PNL) under NRC contract and with advice from EPA. We recommend that the NRC Staff relay these comments to the PNL Staff.

References:

1. Federal Register Notice, Environmental Protection Agency, 40 CFR Part 194, Radiation Protection Criteria for Cleanup of Land and Facilities Contaminated with Residual Radioactive Materials; Advance Notice of Proposed Rulemaking, FR Vol. 51, No. 117, pp.

2264-2266, dated June 18, 1986

2. Letter from ACRS for W. J. Dircks, EDO,

Subject:

ACRS Comments on Proposed Amendments to 10 CFR 20 to specify Residual Radioactive Contamination Limits, dated May 14, 1984

3. Presentation Handout, EPA's Development of Residual Radioactivity Criteria, S. Lichtman, Guides and Criteria Branch, USEPA, dated July 21, 1986 224

l August 8, 1986 Report No. 5 ACRS Waste Management Subcommittee Comments on'the Salvaging of Contaminated Smelted Alloys July 21-23, 1986

1. The Subcommittee continues to believe (Reference 1) that the question of the release of contaminated materials from enrichment plants is but a small part of a larger generic question concerning the disposition of a wide range of related materials, each contaminated by a very small concentration of radioactive materials. We are encouraged by the joint efforts of the NRC and EPA staffs to develop criterfa, guidance and standards relative to the generic implications of this subject (Reference 2).

2. Relative to the matter of the smelted alloys, however, we find the Draft Final Environmental Statement (Reference 3) to be inadequate.

The report does not clearly specify the bases on which the evaluations have been made; it does not adequately support the underlying assumptions; it does not adequately address the decontamination of alloys; and it contains what appear to be many errors and/or incomplete statements and, tables. A key factor is the concentration of uranium somewhat arbitrarily assumed to be present in the various alloys. Another assumption is that dilution of the radionuclides will be adequate to make subsequent (second generation) products acceptable in the public sector. If acceptability is the goal, it could be accomplished in a simpler and more positive manner. The key assumptions leading to the acceptability of the proposed approach are not substantiated. The Subcommittee strongly urges that this document (NUREG-0518) not be published.

3. Although we understand that work on this specific subject has been terminated by the NRC Staff, we understand that DOE is investigating possible alternatives for the release or reuse of smelted alloys. We encourage the NRC Staff to keep abreast of these developments and be prepared to review the DOE plans, if appropriate. If the alloys can be recycled within DOE or D0D operations, but not to the general public, this should be acceptable providing the levels of contamination are reduced as indicated in the report. If the alloys are destined for public use, we believe such action should be carefully reviewed, eyJl uated, and approved by responsible Federal and/or State agencies before being implemented. Considerable added assurance may be necessary to ensure that the meamrements on large batches of alloys are representative, and that concentration mechanisms for unwanted contaminants will not be operative.

225

s SALVAGING OF CONTAMINATED SMELTED ALLOYS 2

References:

1. Letter from ACRS for N. J. Palladino, Chairman, NRC,

Subject:

ACRS Comments on Salvaging of Contaminated Smelted Alloys, dated May 13, 1986

2. ACRS Waste Management Subcommittee Comments on Development of Residual Radiation Limits etc., (Report No. 4), attachment to ACRS letter to V. Stello, EDO,

Subject:

ACRS Comments on Various NMSS and RES Waste Management Topics, dated August 13, 1986.

3. Draft Final Environmental Statement (NUREG-0518), concerning proposed rulemaking exemption from licensing requirements for smelted alloys containing residual technetium-99 and low-enriched uranium, dated February 1984

4. SECY-85-373,

Subject:

Denial of DOE Request for Exemption to Permit Salvaging of Contaminated Smelted Alloys, dated November 25, 1985.

i 226

, . _ -_ . ~ _ _ _ _ _ _ _ . _ _ _ _ - - _ _ _

/

5 >

f August 8, 1986 Report No. 6

^

ACRS Waste Management Subcommittee Comments On NRC Staff Policy Statement And Implementation of NRC Policy on Radioactive Wastes Below Regulatory Concern July 21-23, 1986 l 1. In general, we believe~the decision criteria being developed by the NRC Staff for judging whether to grant a petition for designating radioac-tive wastes as '"below regulatory concern" (BRC), thus pennitting them to be disposed of by conventional means, are good. However, we have the '

following suggestions and coments: <

, 'a. Before the criteria are confinned, it duld be useful to use them i to evaluate a range of potential waste candidates to determine if any would be found to be "below regulatory concern.'" We are concerned that the dose equivalent limit to an individual member of the public~ is so low that successful application of the criteria may prove to be rare.

b. We believe that the models to be used for calculating doses to individual members of the public resulting from the disposal of radioactive wastes should be spedfiedu Included should be a statement relative to the uncertainties acceptable in such models.

i

c. The computer code proposed for use in judging the impact of handling wastes containing radioactive materials in quantities or concentrations "below regulatory concern" is said to be conserva--

tive. The amount of conservatism, however, is unknown. Because of the small dose equivalent that will be acceptable, and because the inclusion of conservatism in modeling will produce unidentifiable distortions of the calculated results, we recomend that the

calculational methods be designed to give best estimate results insofar as is feasible.

d. Although we endorse,the efforts of the NRC Staff to, develop suit-able dose estimation models for evaluation of proposals submitted by petitioners, we believe that greater use might have been made of the methodology described in earlier reports on this general subject prepared by Ford, Bacony Davis, Utah, Inc.

2. The Policy Statement (Reference 1) recomends that evaluations be based j on effective dose equivalents. We suggest that this same

system be used j in conRaring doses from various radiation sources. For example, the

' effective dose equivalent from the natural radiation background should

. include the lung dose contribution from radon. ,

i 227 l ,

Waste Management 2

3. One of the examples cited as providing perspective is the EPA limit for radionuclides in drinking water, which permits members of the public to i receive a maximum dose of 4 mrem per year to an individual organ. This is a dose limit and we believe it is incorrect to cite it as representa- i tive of a dose considered to be "below regulatory concern." j

4. We are encouraged to note the use by the NRC Staff, in the development of this Policy Statement, of several publications of the International Commission on Radiological Protection.

References:

1. SECY-86-204, Policy Issue (Affirmation), Policy Statement on Radioactive Waste Below Regulatory Concern, V. Stello, EDO, to the Commissioners, dated July 11, 1986 l

l l

l l

l 228

Sa ar roq'o UNITED STATES g

8 o NUCLEAR REGULATORY COMMISSION

1. : ADVISORY COMMITTEE ON REACTOR SAFEGUARDS

'* wAsmNGTON, D. C. 20555 i s...+/

o December 16, 1986 1

(

i l

i i

1 j Mr. Victor Stello, Jr.

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

SUBJECT:

ACRS COMMENTS ON THE NRC STAFF REVIEW 0F DOE'S FINAL ENVIRONMENTAL ASSESSMENTS OF HIGH-LEVEL WASTE REPOSITORY SITES On December 4-5, 1986, the Advisory Committee on Reactor Safeguards' Subcommittee on Waste Management met with the NRC Staff to discuss the Staff's review of the U. S. Department of Energy's Final Environmental Assessments of the five candidate sites for a high-level waste reposi-tory. This matter was also discussed during the 320th meeting of the full Committee, December 11-13, 1986.

On the basis of the discussions at these meetings, we believe that the NRC Staff is focusing on the proper items in its review, that the Staff is taking into account the relevant phenomena in conducting its evalua-tions, and that its approach appears reasonable. We also concur that the Staff's primary objective should be to direct its attention to changes that the Department of Energy can incorporate into the Site Characterization Plans (yet to be prepared), rather than seeking addi-tional changes to the existing Final Environmental Assessments.

Sincerely, David A. Ward Chairman l

229

u s auctuu ttrutefoav CoMMissio= i *EPoaY ~uweta <ue.,* no , v, . ,, P, groa= ==

5"3E' BISUOGRAPHIC DATA SHEET NUREG-ll25, olume 8 sEE IN57muCTIOhs 04 THE mEvEm 2 YITLE AND sueTITLE 3 LEAVE SLANK A Compilation o Reports of the Advisory Committee on Rea tor Safeguards. 1986 4 DATE REPORT COMPLETED MoNTM VEAR a suT,,onis, Januap9 1987 f . oATE aEPoaf issuto I f.o,1 . EAR Apr,11 1987 7 FEAponMtNo ORGAN 12 Af aQN N AME AND M AIUNG ado s5 (isscher te Cogyj 8 PROJECTsT ASK.vvoRK umsi NUMBER i \

Advisory Committe on Rea tor Safeguards ../oaoaA~T~uMeta l U.S. Nuclear Regulatory mmission [

Washington, D.C. 20555 /

io sPo~soais.c oacA=,1 Afio= = AME A=o MAiu o Aoo.Ess ,,,,,,,, ,i, C , ii. TvPEo.at,ani Compilation Same as above / D PEntOD COVERED tisscsers,we deress

/

/ Jan. thru Dec. 1986 g /

12 SUPPLEMENT ARY NOTES l

)

i i3 d.05TR ACT flop eeres ersessJ /

/

This compilation contains 58 AGRS/' reports submitted to the Commission or to the Execut Ve Director for Operations i

during the calendar year 1986. 11 reports have been made available to the public through# he NRC Public Document Room and the U.S. Library of Congh.ess. The reports are divided into two groups: Part/l: ADRS Reports on Project Reviews, and Part 2: ACRS Reports dp Generic Subjects.

Part1containsACRSreports'alphabktizedbyproject name and within project nam,e by chrortological order.

Part 2 categorizes the reports by the\most appropriate generic subject area and w'ithin subjedt area by chron-ological order.

/

/

e

/

/

/

/

/

.4 DOCUMENT ANALv515 - a KEvwORDS/DESCRtPTORS 16 AvaiLA9 tut V STATEMENT Nuclear Reactors Safety Engineering Nuclear Reactor Sa'fety Safety Research Unlimited Reactor Operations ,e steunifv CtAssiciCAfio~

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