ML23151A602

PR-050,052,100 - 57FR47802 - Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of PRM-50-20

ML23151A602
Person / Time
Issue date:	10/20/1992
From:	Chilk S NRC/SECY
To:
References
57FR47802, PR-050, PR-052, PR-100
Download: ML23151A602 (1)

Text

ADAMS Template: SECY-067 DOCUMENT DATE: 10/20/1992 TITLE: PR-050, 052, 100 - 57FR47802 - REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PRM-50-20 CASE

REFERENCE:

PR-050, 052, 100 57FR47802 KEYWORD: RULEMAKING COMMENTS Document Sensitivity: Non-sensitive - SUNSI Review Complete

STATUS OF RULEMAKING PROPOSED RULE: PR-050, 052, 100 OPEN ITEM (Y/N) N RULE NAME: REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PRM-50-20 PROPOSED RULE FED REG CITE: 57FR47802 PROPOSED RULE PUBLICATION DATE: 10/20 / 92 NUMBER OF COMMENTS: 83 ORIGINAL DATE FOR COMMENTS: 02/17 / 93 EXTENSION DATE: I I FINAL RULE FED. REG . CITE: 61FR65157 FINAL RULE PUBLICATION DATE: 12 /11/ 96 NOTES ON: WHEN ISSUED AS A FINAL RULE WOULD ALSO DENY PRM-50-20 SUBMITTED BY STATUS : THE FREE ENVIRONMENT, INC., AND PUB. ON 5/19/77 AT 42 FR 25785.

,.. OF RULE : FILE LOCATED ON Pl.

HISTORY OF THE RULE PART AFFECTED: PR-050, 052, 100 RULE TITLE: REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PRM-50-20 PROPOSED RULE PROPOSED RULE DATE PROPOSED RULE SECY PAPER: 92-215 SRM DATE: 08 / 18 / 92 SIGNED BY SECRETARY: 10/13/92 FINAL RULE FINAL RULE DATE FINAL RULE SECY PAPER: 96-118 SRM DATE : 10/11/96 SIGNED BY SECRETARY: 12 / 02 / 96

- STAFF CONTACTS ON THE RULE CONTACTl : DR. ANDREW J. MURPHY MAIL STOP: NL-217A PHONE: 492-3860 CONTACT2: MAIL STOP: PHONE:

DOCKET NO. PR-050, 052, 100 (57FR47802)

In the Matter of REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PRM-50-20 .

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT

- 10/16/92 10/13/92 FEDERAL REGISTER NOTICE - PROPOSED RULE 01/04/93 12/26/92 COMMENT OF SIERRA CLUB - NEW JERSEY CHAPTER (SIDNEY J. GOODMAN) ( 1) 01/06/93 12/30/92 COMMENT OF PAUL MOSS ( 2) 01/07/93 12/29/93 FEDERAL REGISTER NOTICE EXTENDING THE TIME FOR COMMENT ON THE RULE TO 3/24/93. NOTICE PUBLISHED ON 1/5/93 AT 58 FR 271 .

01/11/93 01/07/93 COMMENT OF SAN LUIS OBISPO MOTHERS FOR PEACE (JILL ZAMEK, TREASURER) ( 3) 01/12/93 01/09/93 COMMENT OF DAVID NIXON ( 4)

- 01/15/93 01/08/93 COMMENT OF JOAN 0. KING ( 5) 01/15/93 01/09/93 COMMENT OF TOLEDO COALITION FOR SAFE ENERGY (CHARLENE JOHNSTON) ( 6) 01/19/93 01/13/93 COMMENT OF ALLIANCE FOR SURVIVAL (BARBARA GARTNER, DIRECTOR) ( 7) 01/21/93 12/16/92 LTR FROM JOHN P. RONAFALVY (NUMARC) TO LEONARD SOFFER INQUIRING AS TO WHETHER THE 3/24/93 DUE DATE FOR COMMENTS APPLIES TO PR RELATED DOCUMENTS 01/21/93 01/19/93 COMMENT OF BILL NIERSTEDT ( 8) 02/01/93 01/22/93 COMMENT OF BOB BRISTER ( 9) 02/01/93 01/26/93 COMMENT OF SEACOAST ANTI-POLLUTION LEAGUE (CHARLES W. PRATT) ( 10) 02/02/93 01/27/93 COMMENT OF A. DAVID ROSSIN ( 11) 02/04/93 01/04/93 COMMENT OF J. COURTLAND ROBINSON, M.D., MPH ( 12)

DOCKET NO. PR-050, 052, 100 (57FR47802)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 02/08/93 02/03/93 COMMENT OF JAMES A. MARTIN, JR. ( 13) 02/08/93 02/05/93 COMMENT OF ELIZABETH H. MEIKLEJOHN ( 14) 02/10/93 02/09/93 LETTER FROM CONGRESSMAN C.W. BILL YOUNG TO THE SECRETARY, ENCLOSING THE COMMENT OF 808 BRISTER (COMMENT NUMBER 9) .

02/12/93 02/08/93 COMMENT OF BRUCE CAMPBELL ( 15)

- 02/12/93 01/15/93 LETTER FROM BONNIE BETANCOURT, DEPT. OF ENERGY TO JOSEPH FOUCHARD, OPA, ENCLOSING A COMMENT OF J. COURTLAND ROBINSON, MD (SEE COMMENT NUMBER 12) 02/16/93 02/01/93 COMMENT OF EVA MANSELL ( 16) 02/16/93 02/08/93 COMMENT OF DEIRDRE DONCHIAN ( 17) 02/16/93 02/11/93 COMMENT OF AD HOC COMMITTEE TO REPLACE INDIAN POINT (ANNA MAYO) ( 18) 02/16/93 02/13/93 COMMENT OF JOHN W. G. TUTHILL ( 19) 02/17/93 02/07/93 COMMENT OF DINI SCHUT ( 20) 02/18/93 02/18/93 COMMENT OF REPUBLIC OF CHINA ATOMIC ENERGY COUNCIL (TSING-TUNG HUANG) ( 21)

- 02/19/93 02/14/93 COMMENT OF ANS SPECIAL COMMITTEE ON NEW CONSTRUCTION (EDWARD L. QUINN &KYLE H. TURNER) ( 22) 02/22/93 12/29/92 COMMENT OF DAVID LEISING ( 23) 02/22/93 02/09/93 COMMENT OF GENERAL ATOMICS (R. M. FORSSELL, SR. V.P.) ( 24) 02/23/93 02/12/93 COMMENT OF DR. Z. REYTBLATT ( 25) 02/23/93 02/18/93 COMMENT OF ECOLOGY CENTER OF SOUTHERN CALIFORNIA (ALBERT PINKERSON) ( 26) 02/26/93 12/22/92 COMMENT OF KOREA ELECTRIC POWER CORP (CHUNG, BO HUN, V. P.) ( 27) 03/02/93 02/23/93 COMMENT OF NUCLEAR SAFETY INSTITUTE (AIB-VINCOTTE)

(J. VERLAEKEN &B. DE BOECK) ( 28) 03/02/93 02/26/93 LTR FROM WILLIAM O. DOUB, OF NEWMAN AND HOLTZINGER P.C. REQUESTING ON BEHALF OF THE INTERNATIONAL SITING GROUP AN EXTENSION OF COMMENT PERIOD TO 6/1

DOCKET NO. PR-050, 052, 100 (57FR47802)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 03/05/93 03/01/93 COMMENT OF CORPS OF ENGINEERS (ELLIS L. KRINITZSKY) ( 29) 03/08/93 03/05/93 COMMENT OF ASSOCIATION OF ENGINEERING GEOLOGISTS (JEFFREY R. KEATON, PRESIDENT) ( 30) 03/12/93 03/03/93 COMMENT OF W. SCOTT DUNBAR ( 31) 03/12/93 02/16/93 COMMENT OF OHIO DEPARTMENT OF NATURAL RESOURCES (DR. MICHAEL C. HANSEN) ( 32)

- 03/12/93 01/19/93 COMMENT OF NORTH DAKOTA GEOLOGICAL SURVEY (JOHN P. BLUEMLE) ( 33) 03/15/93 03/15/93 COMMENT OF FEDERATION OF ELECTRIC POWER COMPANIES (RYO IKEGAME, CHAIRMAN) ( 34) 03/17/93 03/12/93 LTR FROM CHAIRMAN SELIN TO RYO IKEGAME, CHAIRMAN NUCLEAR POWER DEVELOPMEMT COUNCIL OF THE FEDERATION OF ELECTRIC POWER COMPANIES, RE COMMENT 03/17 /93 03/10/93 COMMENT OF ELECTRICITE DE FRANCE (REMY CARLE, EXEC. V. P.) ( 35) 03/22/93 03/15/93 COMMENT OF NUCLEAR POWER ENGINEERING CORPORATION (MASAYOSHI SHIBA, DIRECTOR GENERAL) ( 36) 03/22/93 03/18/93 COMMENT OF VEREINIGUNG DEUTSCHER ELEKTRIZTATSWERKE (DR. JOACHIM GRAWE) ( 37) 03/22/93 03/18/93 COMMENT OF NEW YORK POWER AUTHORITY (RALPH E. BEEDLE) ( 38) 03/23/93 03/22/93 COMMENT OF SCOTTISH NUCLEAR LTD (R. J. KILLICK) ( 39) 03/23/93 03/22/93 COMMENT OF G C SLAGIS ASSOCIATES (GERRY C. SLAGIS) ( 40) 03/23/93 03/23/93 COMMENT OF ENEL (INGG. VELONA-FORNACIARI) ( 41) 03/23/93 03/22/93 FEDERAL REGISTER NOTICE EXTENDING THE COMMENT PERIOD TO JUNE 1, 1993. THE NOTICE WAS PUBLISHED ON 3/26/93 AT 58 FR 16377.

03/24/93 03/15/93 COMMENT OF MONTANA BUREAU OF MINES AND GEOLOGY (EDWARD T. RUPPEL, DIRECTOR) ( 42) 03/24/93 03/22/93 COMMENT OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC (SUSAN L. HIATT, DIRECTOR) ( 43) 03/24/93 03/23/93 COMMENT OF YANKEE ATOMIC ELECTRIC CO (D. W. EDWARDS) ( 44)

DOCKET NO. PR-050, 052, 100 (57FR47802)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 03/24/93 03/23/93 COMMENT OF CALIFORNIA DEPARTMENT OF CONSERVATION (JAMES F. DAVIS) ( 45) 03/24/93 03/24/93 COMMENT OF GEORGIA POWER CO (J. T. BECKHAM , JR., V.P.) ( 46) 03/24/93 03/24/93 COMMENT OF SOUTHERN NUCLEAR OPERATING CO (J. D. WOODARD) ( 47) 03/24/93 03/24/93 COMMENT OF NORTHEAST UTILITIES & WASH . PUB . POWER SUPPLY SYS.

(MARK J . WETTERHAHN & K. M. KALOWSKY) ( 56)

- 03/25/93 03/22/93 COMMENT OF VIRGINIA POWER (WILLIAM L. STEWART , SENIOR, V.P.) ( 48) 03/25/93 03/24/93 COMMENT OF ENEA (GIOVANNI NASCHI) ( 49) 03/25/93 03/24/93 COMMENT OF NUCLEAR MANAGEMENT & RESOURCES COUNCIL (WILLIAM H. RASIN, V. P.) ( 50) 03/25/93 03/24/93 COMMENT OF NUCLEAR INFORMATION & RESOURCE SERVICE (NIRS) ( 51) 03/25/93 03/25/93 CORRECTION NOTICE SUBMITTED BY THE OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC ., CORRECTING PAGE 11, PARAGRAPH ONE OF COMMENT NUMBER 43 .

- 03/25/93 03/24/93 COMMENT OF DEPARTMENT OF ENERGY (DWIGHT E. SHELOR) ( 52) 03/25/93 03/24/93 COMMENT OF WESTINGHOUSE ELECTRIC CORP (ENERGY SYS)

(N . J . LIPARULO) ( 53) 03/25/93 03/24/93 COMMENT OF PUBLIC CITIZEN (JAMES P. RICCIO , ESQUIRE) ( 54) 03/26/93 03/24/93 COMMENT OF NIAGARA MOHAWK POWER CORP (C. D. TERRY, V. P.) ( 55) 03/26/93 03/23/93 COMMENT OF GENERAL ELECTRIC CO (P. W. MARRIOTT) ( 57) 03/29/93 03/24/93 COMMENT OF SUSAN BURKE ( 58) 03/29/93 03/23/93 COMMENT OF ENTERGY OPERATIONS , INC (JOHN R. MCGAHA, V. P.) ( 59) 03/29/93 03/23/93 COMMENT OF TWELVE FOREIGN ELECTRIC COMPANIES (JANET E. B. ECKER) ( 60) 03/30/93 03/24/93 COMMENT OF GULF STATES UTILITIES CO (J. E. BOOKER) ( 61)

DOCKET NO. PR-050, 052, 100 (57FR47802)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 03/30/93 03/24/93 COMMENT OF SOUTH CAROLINA ELECTRIC & GAS CO (JOHN L. SKOLDS, V. P.) ( 62) 03/30/93 03/24/93 COMMENT OF FLORIDA POWER & LIGHT CO (W. H. BOHLKE, V.P.) ( 64) 04/01/93 03/31/93 COMMENT OF NUCLEAR ELECTRIC (DR. B. EDMONDSON) ( 63) 04/08/93 03/31/93 COMMENT OF MINISTERE DE L'INDUSTRIE ETC (MICHEL LAVERIE &WALTER HOHLEFELDER) ( 65)

- 04/14/93 03/10/93 COMMENT OF DELAWARE GEOLOGICAL SURVEY (THOMAS E. PICKETT, ASSOC. DIR.) ( 66) 04/26/93 04/22/93 COMMENT OF TENNESSEE VALLEY AUTHORITY (MARK J. BURZYNSKI) ( 67) 05/03/93 04/23/93 COMMENT OF FLORIDA POWER CORP (ROLF C. WIDELL) ( 68) 05/24/93 05/17/93 COMMENT OF DEPARTMENT OF ENERGY (JEFFREY K. KIMBALL) ( 69) 05/26/93 05/26/93 COMMENT OF NATIONAL ATOMIC ENERGY AGENCY (DJALI AHIMSA, DIRECTOR GENERAL) ( 70) 05/28/93 05/28/93 COMMENT OF NUCLEAR MANAGEMENT & RESOURCES COUNCIL (WILLIAM H. RASIN) ( 71)

- 06/01/93 05/28/93 COMMENT OF DEPARTMENT OF ENERGY (E. C. BROLIN,) ( 72) 06/01/93 05/27/93 COMMENT OF NORMAN R. TILFORD ( 73) 06/01/93 06/01/93 COMMENT OF INTERNATIONAL SITING GROUP (WILLIAM O. DOUB, ESQUIRE) ( 74) 06/14/93 06/02/93 COMMENT OF U.S. GEOLOGICAL SURVEY (DALLAS L. PECK, DIRECTOR) ( 75) 06/17/93 03/23/93 COMMENT OF ILLINOIS STATE GEOLOGICAL SURVEY (MORRIS W. LEIGHTON, CHIEF) ( 76) 06/17/93 05/02/93 COMMENT OF ATOMIC ENERGY COMMISSION OF ISRAEL (DR. Y. WEILER) ( 77) 06/28/93 06/24/93 COMMENT OF AMERICAN NUCLEAR SOCIETY (DR. WALTER H. D'ARDENNE) ( 78) 06/29/93 03/23/93 COMMENT OF SARGENT &LUNDY ENGINEERS (B. A. ERLER) ( 79) 06/29/93 03/23/93 COMMENT OF VERMONT AGENCY OF NATURAL RESOURCES (LAURENCE R. BECKER) ( 80)

DOCKET NO . PR-050, 052 , 100 (57FR47802)

DATE DATE OF TITLE OR DOCKETED DOCUMENT DESCRIPTION OF DOCUMENT 06/29/93 03/30/93 COMMENT OF TU ELECTRIC (WILLIAM J. CAHILL , JR., V.P.) ( 81) 06/29/93 04/21/93 COMMENT OF NORTHERN STATES POWER CO (ROGER 0. ANDERSON) ( 82) 07/22/93 07/12/93 COMMENT OF EQE INTERNATIONAL, INC (DR. JAMES J . JOHNSON , PRESIDENT) { 83) 10/04/93 10/01/93 SUPPLEMENT TO COMMENT NO . 74, SUBMITTED BY

- 12/04/96 12/02/96 WILLIAM O. DOUB ( NEWMAN & HOLTZINGER) ON BEHALF OF THE INTERNATIONAL SITING GROUP FEDERAL REGISTER NOTICE - FINAL RULE

DOCKET NLMBER PROPOSED RULE PB 50 -s i <I" / DO 1

( 51 Ft<J.f7<i02)

'96 OE -4 AlO :46 NUCLEAR REGULATORY COMMISSION 10 CFR Parts 21, 50, 52, 54 and 100 0FFI Ct (

DDCl\-1 .

RIN 3150-AD93 Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for 0 Nuclear Power Plants and Denial of Petition from Free Environment, Inc. et. al.

AGENCY : Nuclear Regulatory Co11111ission.

ACTION : Final rule and denial of petition from Free Environment, Inc.

et. a1 *

SUMMARY

The Nuclear Regulatory Co11111ission (NRC) is amending its regulations to update the criteria used in decisions regarding power reactor siting,

~

including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. The rule allows NRC to benefit from experience gained in the application of the procedures and methods set forth in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. This rule primarily consists of two separate changes, namely, the source term and dose considerations, and the seismic and earthquake engineering considerations of reactor siting. The Co11111ission also is denying the remaining issue in petition (PRM-50-20) filed by Free Environment, Inc. et. al.

1/10/q1 EFFECTIVE DATE: (30 days after publication in the Federal Register).

FOR FURTHER INFORMATION CONTACT: Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Co11111ission, Washington, DC 20555-0001, telephone (301) 415-6010, concerning the seismic and earthquake engineering aspects and Mr. Charles E. Ader, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Co11111ission, Washington , DC 20555-0001, telephone (301) 415-5622, concerning other siting aspects.

SUPPLEMENTARY INFORMATION:

I. Background.

II. Objectives.

III. Genesis.

IV. Alternatives.

V. Major Changes.

A. Reactor Siting Criteria (Nonseismic).

B. Seismic and Earthquake Engineering Criteria.

VI. Related Regulatory Guides and Standard Review Plan Sections.

VII. Future Regulatory Action.

VIII. Referenced Documents.

IX. Su11111ary of Co11111ents on the Proposed Regulations.

A. Reactor Siting Criteria (Nonseismic).

B. Seismic and Earthquake Engineering Criteria.

X. Small Business Regulatory Enforcement Fairness Act XI. Finding of No Significant Environmental Impact: Availability.

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XII. Paperwork Reduction Act Statement.

XIII. Regulatory Analysis.

XIV. Regulatory Flexibility Certification.

XV. Backfit Analysis.

I. Background The present regulation regarding reactor site criteria (10 CFR Part 100) was pr011ulgated April 12, 1962 (27 FR 3509). NRC staff guidance on exclusion area and low population zone sizes as well as population density was issued in Regulatory Guide 4.7, *General Site Suitability Criteria for Nuclear Power Stations," published for coment in September 1974. Revision 1 to this guide was issued in November 1975. On June 1, 1976, the Public Interest Research Group (PIRG) filed a petition for rulemaking (PRM-100-2) requesting that the NRC incorporate minimum exclusion area and low population zone distances and population density limits into the regulations. On April 28, 1977, Free Environment, Inc. et. al., filed a petition for rulemaking (PRM-50-20). The remaining issue of this petition requests that the central Iowa nuclear project and other reactors be sited at least 40 miles from major population centers. In August 1978, the Connission directed the NRC staff to develop a general policy statement on nuclear power reactor siting. The *Report of the Siting Policy Task Force* (NUREG--0625) was issued in August 1979 and provided recormiendations regarding siting of future nuclear power reactors. In the 1980 Authorization Act for the NRC, the Congress directed the NRC to decouple siting from design and to specify demographic criteria for siting. On July 29, 1980 (45 FR 50350), the NRC issued an Advance Notice of Proposed 3

Rulemaking (ANPRM} regarding revision of the reactor site criteria, which discussed the reco111111endations of the Siting Policy Task Force and sought public connents. The proposed rulemaking was deferred by the Co11111ission in December 1981 to await development of a Safety Goal and improved research on accident source terms. On August 4, 1986 (51 FR 23044), the NRC issued its Policy Statement on Safety Goals that stated quantitative health objectives with regard to both prompt and latent cancer fatality risks. On December 14, 1988 (53 FR 50232), the NRC denied PRH-100-2 on the basis that it would unnecessarily restrict NRC's regulatory siting policies and would not result in a substantial increase in the overall protection of the public health and safety. The Co11111ission is addressing the remaining issue in PRM-50-20 as part of this rulemaking action.

Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants,* to 10 CFR Part 100 was originally issued as a proposed regulation on Novelllber 25, 1971 (36 FR 22601), published as a final regulation on November 13, 1973 (38 FR 31279), and became effective on December 13, 1973. There have been two amendments to 10 CFR Part 100, Appendix A. The fir~t amendment, issued November 27, 1973 (38 FR 32575), corrected the final regulation by adding the legend under the diagram. The second amendment resulted from a petition for rulemaking (PRH 100-1) requesting that an opinion be issued that would interpret and clarify Appendix A with respect to the determination of the Safe Shutdown Earthquake. A notice of filing of the petition was published on Hay 14, 1975 (40 FR 20983). The substance of the petitioner's proposal was accepted and published as an ill'lllediately effective final regulation on January 10, 1977 (42 FR 2052).

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The first proposed revision to these regulations was published for public connent on October 20, 1992, {57 FR 47802). The availability of the five draft regulatory guides and the standard review plan section that were developed to provide guidance on meeting the proposed regulations was published on November 25, 1992, (57 FR 55601). The comment period for the proposed regulations was extended two times. Firstt the NRC staff initiated an extension (58 FR 271; January 5, 1993) ftom February 17, 1993 to March 24, 1993, to be consistent with the comment period on the draft regulatory guides and standard review plan section. Second, in response to a request from the public, the c0f1lllent period was extended to June 1, 1993 (58 FR 16377; March 26, 1993).

The second proposed revision to these regulations was published for public conoent on October 17, 1994 (59 FR 52255). The NRC stated on February 8, 1995, (60 FR 7467) that it intended to extend the convnent period to allow interested persons adequate time to provide c011111ents on staff guidance documents. On February 28, 1995, the availability of the five draft regulatory guides *and three standard review plan sections that were developed to provide guidance on meeting the proposed regulations was published (60 fR 10880) and the connent period for the proposed rule was extended to May 12, 1995 (60 FR 10810).

II. Objectives The objectives of this regulatory action are to --

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1. State basic site criteria for future sites that, based upon experience and importance to risk, have been shown as key to protecting public health and safety;

2. Provide a stable regulatory basis for seismic and geologic siting and applicable earthquake engineering design of future nuclear power plants that will update and clarify regulatory requirements and provide a flexible structure to pemit consideration of new technical understandings; and

3. Relocate source term and dose requirements that apply primarily to plant design into 10 CFR Part 50.

III. Genesis The regulatory action reflects changes that are intended to (1) benefit from the experience gained in applying the existing regulation and from research; (2) resolve interpretive questions; {3) provide needed regulatory flexibility to incorporate state-of-the-art improvements in the geosciences and earthquake engineering; and (4) simplify the language to a more *plain English" text.

The new requirements in this rulemaking apply to applicants who appl~

for a construction permit, operating license, preliminary design approval, final design approval, manufacturing license, early site permit, design certification, or combined license on or after the effective date of the final regulations. However, for those operating license applicants and holders whose construction permits were issued prior to the effective date of this 6

final regulation, the reactor site criteria in 10 CFR Part 100, and the seismic and geologic siting criteria and the earthquake engineering criteria in Appendix A to 10 CFR Part 100 would continue-to apply in all subsequent proceedings, including license amendnlents and renewal of operating licenses pursuant to 10 CFR Part 54.

Criteria not associated with the selection of the site or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed in 10 CFR Part 50. This action is consistent with the location of other design requirements in 10 CFR Part 50.

Because the revised criteria presented in this final regulation does not apply to existing plants, the licensing bases for existing nuclear power plants must remain a part of the regulations. Therefore, the non-seismic and seismic reactor site criteria for current plants is retained as Subpart A and Appendix A to 10 CFR Part 100, respectively. The revised reactor site criteria is added as Subpart Bin 10 CFR Part 100 and applies to site applications received on or after the effective date of the final regulations.

Non-seismic site criteria is added as a new sl00.21 to Subpart Bin 10 CFR Part 100. The criteria on seismic and geologic siting is added as a new sl00.23 to Subpart Bin 10 CFR Part 100. The dose calculations and the earthquake engineering criteria is located in 10 CFR Part 50 (s50.34(a) and Appendix S, respectively). Because Appendix Sis not self executing, applicable sections of Part 50 (sS0.34 and s50.54) are revised to reference

• Appendix S. The regulation also makes crnforming amendments to 10 CFR Parts 21, 50, 52, and 54. Sections 21.3, 50.49(b)(l), 50.65(b)(l), 52.17(a)(l), and 54.4(a)(1)(1ii) are amended to reflect changes ins 50.34(a)(l) and 10 CFR Part 100.

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. IV. Alternatives The first alternative considered by the COD1Rission was to continue using current regulations for site suitability determinations. This is not considered an acceptable alternative. Accident source terms and dose calculations currently primarily influence plant design requirements rather than siting. It is desirable to state basic site criteria which, through importance to risk, have been shown to be key to assuring public health and safety. Further, signifi'cant advances in understanding severe accident behavior, including fission product release and transport, as well as in the*

earth sciences and in earthquake engineering have taken place since the promulgation of the present regulation and deserve to be reflected in the regulations.

The second alternative considered was replacement of the existing regulation with an entirely new regulation. This is not an acceptable alternative because the provisions of the existing regulations fonn part of the licensing bases for many of the operating nuclear power plants and.others that are in various stages of obtaining operating licenses. Therefore, these provisions should remain in force and effect.

The approach of establishing the revised requirements in new sections to 10 CFR Part 100 and relocating plant design requirements to 10 CFR Part 50 while retaining the existing regulation was chosen as the best alternative.

The public will benefit from a clearer, more uniform, and more consistent licensing process that incorporates updated information and is subject to 8

fewer interpretations. The NRC staff will benefit from improved regulatory implementation (both technical and legal), fewer interpretive debates, and increased regulatory flexibility. Applicants will derive the same benefits in addition to avoiding licensing delays caused by unclear regulatory requireinents.

V. MAJOR CHANGES A. R~actor Siting Criteria (Nonseis*ic).

Since promulgation of the reactor site criteria in 1962, the Colllllission has approved more than 75 sites for nuclear power reactors and has had an opportunity to review a number of others. In addition, light-water commercial power reactors have accumulated about 2000 reactor-years of operating experience in the United States. As a result of these site reviews and operational experience, a great deal of insight has been gained regarding the design and operation of nuclear power plants as well as the site factors that influence risk. In addition, an extensive research effort has been conducted to understand accident phenomena, including fission product,.

release and transport. This extensive operational experience together with the insights gained from recent severe accident research as well as numerous risk studies on radioactive material releases to the environment under severe accident conditions have all confirmed that present colllllercial power reactor design, construction, operation and siting is expected to effectively limit risk to the public to very low levels. These risk studies include the early "Reactor 9

Safety Study* (WASH-1400), published in 1975, many Probabilistic Risk Assessinent (PRA) studies conducted on individual plants as well as several specialized studies, and the recent ~Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,* (NUREG-1150), issued in 1990. Advanced reactor designs currently under review are expected to result in even lower risk and i111proved safety compared to existing plants. Hence, the substantial base of knowledge regarding power reactor siting, design, construction and operation reflects that the primary factors that determine public health and safety are the reactor design, construction and operation.

Siting factors and criteria, however, are important in ~ssuring that radiological doses from nonnal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, that site characteristics are such that adequate security*measures to protect the plant can be developed, and that physical characteristics unique to the proposed site that could pose a significant impediment to the development of emergency plans are identified. The Conrnission has also had a long standing policy of siting reactors away from densely populated centers, and 1s continuing this policy in this rule.

The Connission is incorporating basic reactor site criteria in this rule to accomplish the above purposes. The Comission is retaining source tenn and dose calculations to verify the adequacy of a site for a specific plant, but source term and dose calculations are relocated to Part 50, since experience has shown that these calculations have tended to influence plant design aspects such as containment leak rate or filter perfonnance rather than siting. No specific source tennis referenced in Part 50. Rather, the source 10

term is required to be one that is* ... assu1Red to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.* Hence, this guidance can be utilized with the source term currently used for light-water reactors, or used in conjunction with revised accident source tems.

The relocation of source term and dose calculations to Part 50 represent a partial decoupling of siting from accident source term and dose calculations. The siting criteria are envisioned to be utilized together with standardized plant designs whose features will be certified in a separate design certification rulemaking procedure. Each of the standardized designs will specify an atmospheric dilution factor that would be required to be met, in order to meet the dose criteria at the exclusion area boundary. For a given standardized design, a site having relatively poor dispersion characteristics would require a larger exclusion area distance than one having good dispersion characteristics. Additional design features would be discouraged in a standardized design to c011pensate for otherwise poor site conditions.

Although individual plant tradeoffs will be discouraged for a given standardized design, a different standardized design could require a different atmospheric dilution factor. For custom plants that do not involve a standardized design, the source term and dose criteria will continue to provide assurance that the site is acceptable for the proposed design.

Rationale for Individual Criteria A. Exclusion Area. An exclusion area surrounding the imediate vicinity of the plant has been a requirement for siting power reactors from the very 11

beginning. This area provides a high degree of protection to the public from a variety of potential plant accidents and also affords protection to the plant fr0111 potential man-related hazards. The Comission considers an exclusion area to be an essential feature of a reactor site and is retaining this requirement, in Part Su, to verify that an applicant's proposed exclusion area distance is adequate to assure that the radiological dose to an individual will be acceptably low in the event of a postulated accident.

However, as noted above, if source term and dose calculations are used in conjunction with standardized designs, unlimited plant tradeoffs to compensate for poor site conditions will not be permitted. For plants that do not involve standardized designs, the source term and dose calculations will provide assurance that.the site is acceptable for the proposed design.

T~e present regulation requires that the exclusion area be of such size that an individual located at any point on its boundary for two hours innediately following onset of the postulated fission product release would not receive a total radiation dose in excess of 25 rem to the whole body or 300 re to the thyroid gland. A footnote in the present reg11lation notes that a whole body dose of 25 rem has been stated to correspond numerically to the once in a lifetime accidental or emergency dose to radiation workers which could be disregarded in the determination of their radiation exposure status

{NBS Handbook 69 dated June 5, 1959). However, the same footnote also clearly states that the Comission's use of this value does not imply that it considers it to be an acceptable limit for an* emergency dose to the public under accident conditions, but only that it represents a reference value to be used for evaluating plant features and site characteristics intended to mitigate the radiological consequences of accidents 1n order to provide 12

assurance of low risk to the public under postulated accidents. The Connission, based upon extensive experience in applying this criterion, and in recognition of the conservatism of the assumptions in its application (a large fission product release within containment associated with major core damage, maximum allowable containment leak rate, a postulated single failure of any of the fission product cleanup systems, such as the containment sprays, adverse site 1Reteorological dispersion characteristics, an individual presumed to be located at the boundary of the exclusion area at the centerline of the plume for two hours without protective actions), believes that this criterion has clearly resulted in an adequate level of protection. As an illustration of the conservatism of this assessment, the maximum whole body dose received by an actual individual during the Three Mile Island accident in March 1979, which involved Major core damage, was esti~ated to be about 0.1 r,em.

The proposed rule considered two changes in this area.

First, the Commission proposed that the use of different doses for the whole body and thyroid gland be replaced by a single value of 25 rem, total effective dose equivalent {TEDE).

The proposed use of the total effective dose equivalent, or TEDE, was noted as being consistent with Part 20 of the Comission's regulations and was also based upon two considerations. First, since it utilizes a risk consistent methodology to assess the radiological impact of all relevant nuclides upon all body organs, use of TEDE promotes a uniformity and consistency in assessing radiation risk that may not exist with the separate whole body and thyroid organ dose values in the present regulation. Second, use of TEDE lends itself readily to the application of updated accident source terms, which can vary not only with plant design, but in which* additional 13

nuclides, besides the noble gases and iodine are predicted to be released into contain11ent.

The COR111ission considered the current dose criteria of 25 rem whole body and 300 rem thyroid with the intent of selecting a TEDE numerical value equivalent to the risk implied by the current dose criteria. The Commission proposed to use the risk of latent cancer fatality as the appropriate risk measure since quantitative health objectives (QHOs) for it have been established in the Commission's Safety Goal policy. Although the supplefAentary inforR1ation in the proposed rule noted that the current dose criteria are equivalent in risk to 27 rem TEDE, the Commission proposed to use 25 rem TEDE as the dose criterion for plant evaluation purposes, since this value is essentially the same level of risk as the current criteria.

However, the C011111ission specifically requested conments on whether the current dose criteria should be modified to utilize the total effective dose equivalent or TEDE concept, whether a TEDE value of 25 rem (consistent with latent cancer fatality), or 34 rem (consistent with latent cancer incidence),

or some other value should be used, and whether the dose criterion should also include a *capping* limitation, that is, an additional requirement that the dose to any individual organ not be in excess of some fraction of the total.

Based on the co11111ents received, there was a general consensus that the use of the TEDE concept was appropriate, and a nearly unanimous opinion that no organ "capping* dose was required, since the TEDE concept provided the appropriate risk weighting for all body organs.

With regard to the value to be used as the dose criterion, a number of comments were received that the proposed value of 25 rem TEDE represented a more restrictive criterion than the current values of 25 rem whole body and 14

300 rem to the thyroid gland. These connenters noted that the use of organ weighting factors of 1 for the whole body and 0.03 for the thyroid as given in 10 CFR Part 20, would yield a value of 34 rem TEDE for whole body and thyroid doses of 25 and 300 rem, respectively. This is because the organ weighting factors in 10 CFR Part 20 include other effects {e.g., genetic) in addition to latent cancer fatality.

After careful consideration, the Connission has decided to adopt a value of 25 re11 TEDE as the dose acceptance criterion for the final rule. The bases for this decision follows. First, the Connission has generally based its regulations on the risk of latent cancer fatality. Although a numerical calculation would lead to a value of 27 rem TEDE, as noted in the discussion that accompanied the proposed rule, the Commission concludes that a value of 25 rem is sufficiently close, and that the use of 27 rather than-25 implies an unwarranted numerical precision. In addition, in terms of occupational dose, Part 20 also per111its a once-in-a-lifetime planned special dose of 25 rem TEDE.

In addition, EPA guidance sets a limit of 25 rem TEDE for workers performing emergency service such as lifesaving or protection of large populations.

While the COR111ission does not, as noted above, regard this dose value as one that is acceptable for members of the public under accident conditions, it provides a useful perspective with regard to doses that ought not to be exceeded, even for radiation workers under eJRergency conditions.

The argument that a criterion of 25 rem TEDE in conjunction with the organ weighting factors of 10 CFR Part 20 for its calculation represents a tightening of the dose criterion, while true in theory, is not true in practice. A review of the dose analyses for operating plants has shown that the thyroid dose limit of 300 rem has been the limiting dose criterion in 15

licensing reviews, and that all operating plants would be able to meet a dose criterion of 25 rewi TEDE. Hence, the Coanission concludes that, in practice, use of the organ weighting factors of Part 20 together with a dose criterion of 25 re TEOE, represents a relaxation rather than a tightening of the dose criterion. In adopting this value, the Comission also rejects the view, advanced by some, that the dose calculation is merely a *reference" val~e that bears no relation to what might be experienced by an actual person in an accident. Although the Co11111ission considers it highly unlikely that an actual person would receive such a dose, because of the conservative and stylized assumptions employed in its calculation, it is conceivable.

The second change proposed in this area was in regard to the time period that a hypothetical individual is assumed to be at the exclusion area

• boundary. While the duration of the time period remains at a value of two hours, the proposed rule stated that this time period not be fixed in regard to the appearance of fission products within containment, but that various two-hour periods be examined with the objective that the dose to an individual not be in excess of 25 rem TEDE for any two-hour period after the appearance of fission products'within containment. The Comission proposed this change to reflect improved understanding of fission product release into the containment under severe accident conditions. For an assumed instantaneous release of fission products, as contemplated by the present rule, the two hour period that cOAlllences with the onset of the fission product release clearly results in the highest dose to an individual offsite. Improved understanding of severe accidents shows that fission product releases to the containment do not occur instantaneously, and that the bulk of the releases may not take place for about an hour or more. Hence, the two-hour period comencing with the onset 16

of fission product release may not represent the highest dose that an individual could be exposed to over any two-hour period. As a result, the C01111ission proposed that various two-hour periods be examined to assure that the dose to a hypothetical individual at the exclusion area boundary would not be in excess of 25 rem TEDE over any two-hour period after the onset of fission product release.

A nlllllber of c0111ents received in regard to this proposed criterion stated that so-called *sliding* two-hour window for dose evaluation at the exclusion area boundary was confusing, illogical, and inappropriate. Several coamenters felt it was difficult to ascertain which two hour period represented the maximum. Others expressed the view that the significance of such a calculation was not clearly stated nor understood. For example, one C011118nt expressed the view that a dose evaluated for a "slidinga,two-hour period was logically inconsistent since it implied either that an individual was not at the exclusion area boundary prior to the accident, and approached close to the plant after initiation of the accident, contrary to what might be expected, or that the individual was, in fact, located at the exclusion area boundary all along, in which case the dose contribution received prior to the Hmaxi11t1m* two hour value was being ignored.

Although the COl'llfflission recognizes that evaluation of the dose to a hypothetical individual over any two-hour period may not be entirely consistent with the actions of an actual individual in an accident, the intent is to assure that the short-term dose to an individual will not be in excess of the acceptable value, even where there is some variability in the time that an individual might be located at the exclusion area boundary. In addition, the dose calculation should not be taken too literally with regard to the 17

actions of a real individual, but rather is intended primarily as a means to evaluate the effectiveness of the plant design and site characteristics in itigating postulated accidents.

For these reasons, the C01111ission is retaining the requirement, in the final rule, that the dose to an individual located at the nearest exclusion area boundary over any two-hour period after the appearance of fission products in containment, should not be in excess of 25 rem total effective dose equivalent (TEDE).

B. Site Dispersion Factors Site dispersion factors have been utilized to provide an assessment of dose to an individual as a result of a postulated accident. Since the Commission is requiring that a verification be made that the exclusion area distance is adequate to assure that the guideline dose to a hypothetical individual will not be exceeded under postulated accident conditions, as well as to assure that radiological limits are met under normal operating conditions, the Commission is requiring that the atmospheric dispersion characteristics of the site be evaluated, and that site dispersion factors based upon this evaluation be determined and used in assessing radiological consequences of normal operations as well as accidents.

C. Low Population Zone. The present regulation requires that a low population zone (LPZ) be defined illl1l8diately beyond the exclusion area.

Residents are permitted in this area, but the number and density must be such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. In addition, the nearest densely populated center containing more than about 18

25,000 residents must be located no closer than one and one-third times the outer boundary of the LPZ. Finally, the dose to a hypothetical individual located at the outer boundary of the LPZ over the entire course of the accident 1RUst not be in excess of the dose values given in the regulation.

While the Comission considers that the siting functions intended for the LPZ, namely, a low density of residents and the feasibility of taking protective actions, have been accomplished by other regulations or can be accomplished by other guidance, the Cormiission continues to believe that a requirement that limits the radiological consequences over the course of the accident provides a useful evaluation of the plant's long-term capability to mitigate postulated accidents. For this reason, the Convnission is retaining the requirement that the dose consequences be evaluated at the outer boundary of the LPZ over the course of the postulated accident and that these.not be in excess of 25 rem TEDE.

D. Physical Characteristics of the Site It has been required that physical characteristics of the site, such as the geology, seismology, hydrology, meteorology characteristics be considered in the design and construction of any plant proposed to be located there. The final rule requires that these characteristics be evaluated and that site parameters, such as design basis flood conditions or tornado wind loadings be established for use in evaluating any plant to be located on that site in order to ensure that the occurrence of such physical phenomena would pose no undue hazard.

E. Nearby Transportation Routes. Industrial and Military Facilities As for natural phenomena, it has been a long-standing NRC staff practice to 19

review man-related activities in the site vicinity to provide assurance that potential hazards associated with such facilities or transportation routes will pose no undue risk to any plant proposed to be located at the site. The final rule codifies this practice.

F. Adequacy of Security Plans The rule requires that the characteristics of the site be such that adequate security plans and measures for the plant could be developed. The Comission envisfons that this will entail a small secure area considerably smaller than that envisioned for the exclusion area.

G. Emergency Planning The propos~d rule stated that the site characteristics should be such that adequate plans to carry out protective measures for embers of the public in the event of emergency could be developed. To avoid any misinterpretation that the Comission is adopting emergency planning standards that implicitly overrule or may be in conflict with previous Comission decisions (e.g., CLI-90-02}, the lanquage in the final rule has been modified to be consistent with that of section 52.17 of the Connission's regulations regarding early site permits.

The Connission's decision in Seabrook on emergency planning, made in connection with an operating license review for a site previously approved, is being extended in considering site suitability for future reactor sites. The Comission, in its Seabrook decision, CLI-90-02, reiterated its earlier determination in the Shoreham decision, CLI-86-13, that the adequacy of an emergency plan is to be determined by the sixteen planning standards of 10 CFR 50.47(b}, and that these standards do not require that an adequate plan 20

achieve a preset minimum radiation dose saving or a minimum evacuation time for the plume exposure pathway emergency planning zone in the event of a serious accident. Rather, the Commission noted that emergency planning is required as a matter of prudence and for defense-in-depth, and that the adequacy of an emergency plan was to be judged on the basis of its meeting the 16 planning standards given in 10 CFR 50.47(b). Hence, the characteristics of the site, which determine the evacuation time for the plume exposure pathway emergency planning zone, have not entered into the determination of the

- adequacy of an emergency plan. Emergency plans developed according to the above planning standards will result in reasonable assurance that adequate protective measures can be taken in the event of emergency.

It is sufficient that an applicant identify any physical site characteristics that could represent a significant impediment to* the development of emergency plans, primarily to assure that "A range of protective actions have been developed for the plume exposure pathway e1Rergency planning zone for emergency workers and the public", as stated in the planning standards.

Accordingly, appropriate sections of the rule (e.g., s100.2l(g)) have been modified to state that "physical characteristics unique to the proposed site that could pose a significant impediment to the development of emergency plans must be identified." Except for the deletion of the phrase "such as egress liaitations from the area surrounding the site*, this language is identical to that in s52.17(b)(l). This phrase is being deleted from s100.21(g) (but s52.17(b)(l) remains unchanged), to eliminate any confusion that might arise regarding its scope.

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H. Siting Away From Densely Populated centers Population density considerations beyond the exclusion area have been required since issuance of Part 100 in 1962. The current rule requires a *1ow population zone" {LPZ) beyond the imediate exclusion area. The LPZ boundary aust be of such a size that an individual located at its outer boundary must not receive a dose in excess of the values giv~n in Part 100 over the.course of the accident. While numerical values of population or population density are not specified for this region, the regulation also requires that the nearest boundary of a densely populated center of about 25,000 or more persons be located no closer than one and one-third times the LPZ outer boundary.

Part 100 has no population criteria.other than the size of the LPZ and the proximity of the nearest population center, but notes that *where very* large cities are involved, a greater distance may be necessary."

Whereas the exclusion area size is based upon limitation of individual risk, population density requirements serve to set societal risk limitations and reflect consideration of accidents beyond the design basis, or severe accidents. Such accidents were clearly a consideration in the original issuance of Part 100, since the Statement of Considerations {27 FR 3509; April 12, 1962) noted that:

"Further, since accidents of greater potential hazard than those commonly postulated as rep~esenting an upper limit are conceivable, although highly improbable, it was considered desirable to provide for protection against excessive exposure doses to people in large centers, 22

where effective protective measures might not be feasible .*. Hence, the population center distance was added as a site requirement."

Limitation of population density beyond the exclusion area has the following benefits:

(a) It facilitates emergency preparedness and planning; and (b) It reduces potential doses to large numbers of people and reduces property damage in the event of severe accidents.

Although the Conmission's Safety Goal policy provides guidance on*

individual risk limitations, in the form of the Quantitative Health Objectives (QHO), it provides no guidance with regard to societal risk limitations and therefore cannot be used to ascertain whether a particular population density would meet the Safety Goal.

However, results of severe accident risk studies, particularly those obtained from NUREG-1150, can provide useful insights for consideri.ng potential criteria for population density. Severe accidents having the highest consequences are those where core-melt together with early bypass of or containment failure occurs. Such an event would likely lead to a "large releasea (without defining this precisely). Based upon NUREG-1150, the probability of a core-melt accident together with early containment failure or bypass for some current generation LWRs is estimated to be between 10-a and 10~ per reactor year. For future plants, this value is expected to be less than 10-' per reactor year.

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If a reactor was located nearer to a large city than current NRC practice pennitted, the likelihood of exposing a large number of people to significant releases of radioactive material would be about the same as the probability of a core-melt and early containment failure, that is, less than 10-* per reactor year for future reactor designs. It is worth noting that events having the very low likelihood of about IO-' per reactor year or lower have been regarded in past licensing actions to be *incredible", and as such, have not been required to be incorporated into the design basis of the plant.

Hence, based solely upon accident likelihood, it might be argued that siting a reactor nearer to a large city than current NRC practice would pose no undue risk.

If, however, a reactor were sited away from large cities, the likelihood of the city being affected would be reduced because of two factors. First, the likelihood that radioactive material would actually be carried towards the city is reduced because it is likely that the wind will blow in a direction away from the city. Second, the radiological dose consequences would also be reduced with distance because the radioactive material becomes increasingly diluted by the atmosphere and the inventory becomes depleted due to the natural processes of fallout and rainout before reaching the city. Analyses indicate that if a reactor were located at distances ranging from 10 to about 20 miles away from a city, depending upon its size, the likelihood of exposure of large numbers of people within the city would be reduced by factors of ten to one hundred or more compared with locating a reactor very close to a city.

In summary, next-generation reactors are expected to have risk characteristics sufficiently low that the safety of the public is reasonably assured by the reactor and plant design and operation itself, resulting in a 24

very low likelihood of occurrence of a severe accident. Such a plant can satisfy the QHOs of the Safety Goal with a very small exclusion area distance (as low as 0.1 miles). The consequences of design basis accidents, analyzed using revised source terms and with a realistic evaluation of engineered safety features, are likely to be found acceptable at distances of 0.25 miles or less. With regard to population density beyond the exclusion area, siting a reactor closer to a densely populated city than is current NRC practice would pose a very low risk to the populace.

Nevertheless, the Co11111ission concludes that defense-in-depth considerations and the additional enhancement in safety to be gained by siting reactors away from densely populated centers should be maintained.

The C011111ission is incorporating a two-tier approach with regard to population density and reactor sites. The rule requires that reactor sites be located away from very densely populated centers, and that areas of low population density are, generally, preferred. The Comission believes that a site not falling within these two categories, although not preferred, can be found acceptable under certain conditions.

The Commission is not establishing specific numerical criteria for evaluation of population density in siting future reactor facilities because the acceptability of a specific site front the standpoint of population density must be considered in the overall context of safety and environmental considerations. The Comission's intent is to assure that a site that has significant safety, environmental or economic advantages is not rejected solely because it has a higher population density than other available sites.

Population density is but one factor that must be balanced against the other advantages and disadvantages of a particular site in determining the site's 25

acceptability. Thus, it l'AUst be recognized that sites with higher population density, so long as they are located away from very densely populated centers, can be approved by the Co11111ission if they present advantages in terms of other considerations applicable to the evaluation of proposed sites.

Petition Filed By Free Environment, Inc. et. al.

On April 28, 1977, Free Environment, Inc. et. al., filed a petition for rulemaking (PRM-50-20) requesting, among other things, that *the central Iowa nuclear project and other reactors be sited at least 40 miles from major population centers.A The petitioner also stated that "locating reactors in sparsely-populated areas *** has been endorsed in non-binding NRC guidelines for reactor siting.* The petitioner did not specify what constituted a major population center. The only NRC guidelines concerning population density in regard to reactor siting are in Regulatory Guide 4.7, issued in 1974, and revised in 1975, prior to the date of the petition. This guide states population density values of 500 persons per square mile out to a distance of 30 miles fr0111 the reactor, not 40 miles.

Regulatory Guide 4.7 does provide effective separation from population centers of various sizes. Under this guide, a population center of about 25,000 or more residents should be no closer than 4 miles (6.4 km) from a reactor because a density of 500 persons per square mile within this distance would yield a total population of about 25,000 persons. Similarly, a city of 100,000 or more residents should be no closer than about 10 miles (16 km); a city of 500,000 or more persons should be no closer than about 20 miles (32 km), and a city of 1,000,000 or more persons should be no closer than about 30 miles (50 km) from the reactor.

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The Conmission has examined these guidelines with regard to the Safety Goal. The Safety Goal quantitative health objective in regard to latent cancer fatality states that, within a distance of ten miles (16 km) from the reactor, the risk to the population of latent cancer fatality from nuclear power plant operation, including accidents, should not exceed one-tenth of one percent of the likelihood of latent cancer fatalities from all other causes.

In addition to the risks of latent cancer fatalities, the C011111ission has*also investigated the likelihood and extent of land contamination arising from the release of long-lived radioactive species, such as cesium--137, in the event of a severe reactor accident.

The results of these analyses indicate that the latent cancer fatality quantitative health objective noted is met for current plant designs. *From analysis done in support of this proposed change in regulation, the likelihood of permanent relocation of people located 110re than about 20 miles (32 km) from the reactor as a result of land contamination from a severe accident is very low. A revision of Regulatory Guide 4.7 which incorporated this finding that population density guidance beyond 20 miles was not needed in the evaluation of potential reactor sites was issued for conrnent at the time of the proposed rule. No coments were received on this aspect of the guide.

Therefore, the Commission concludes that the NRC staff guidance in Regulatory Guide 4.7 provide a means of locating reactors away from population centers, including *major" population centers, depending upon their size, that would limit societal consequences significantly, in the event of a severe accident. The Comission finds that granting of the petitioner's request to specify population criteria out to 40 miles would not substantially reduce the risks to the public. As noted, the Conmission also believes that a higher 27

population density site could be found to be acceptable, compared to a lower population density site, provided there were safety, environmental, or economic advantages to the higher population site. Granting of the petitioner's request would neglect this possibility and would make population density the sole criterion of site acceptability. For these reasons, the Conmission has decided not to adopt the proposal by Free Environment, Incorporated.

The Comission also notes that future population growth around a nuclear power plant site, as in other areas of the region, is expected but cannot be predicted with great accuracy, particularly in the long-term. Population growth in the site vicinity will be periodically factored into the emergency plan for the site, but since higher population density sites are not unacceptable, per se, the Commission does not intend to cons1der license conditions ~r restrictions upon an operating reactor solely upon the basis that the population density around it may reach or exceed levels that were not expected at the time of site approval. Finally, the Conrnission wishes to emphasize that population considerations as well as other sit1ng requirements apply only for the initial siting for new plants and will not be used in evaluating applications for the renewal of existing nuclear power plant licenses.

Change to 10 CFR Part 50 The change to 10 CFR Part 50 relocates from 10 CFR Part 100 the dose requirements for each applicant at specified distances. Because these requirements affect reactor design rather than siting, they are more appropriately located in 10 CFR Part 50.

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These requirements apply to future applicants for a construction permit, design certification, or an operating license. The Co11111ission will consider, after further experience in the review of certified designs whether more specific requirements need to be developed regarding revised accident source terms and severe accident insights.

8. Seisaic and Earthquake Engineering Criteria.

The following major changes to Appendix A, "Seismic and Geologic Siting Criteria for Nuclear Power Plants,* to 10 CFR Part 100, are associated with the seismic and earthquake engineering criteria rulemaking. These changes reflect new infonnation and research.results, and incorporate the intentions of this regulatory action as defined in Section III of this rule. Much of the following*dtscussion remains unchanged from that issued for public conment (59 FR 52255) because there were no cOnlments which necessitated a major change to the regulations and supporting documentation.

1. Separate Siting from Design.

Criteria not associated with site suitability or establishment of the Safe Shutdown Earthquake Ground Motion {SSE)' have been placed into 10 CFR Part

50. This action is consistent with the location of other design requirements in 10 CFR Part 50. Because the revised criteria presented in the regulation will not be applied to existing plants, the licensing basis for existing nuclear power plants must remain part of the regulations. The criteria on seismic and geologic siting would be designated as a news 100.23 to Subpart B 29

in 10 CFR Part 100. Criteria on earthquake engineering would be designated as a new Appendix S, *Earthquake Engineering Criteria for Nuclear Power Plants,* to 10 CFR Part 50.

2. Remove Detailed Guidance from the Regulation.

Appendix A to 10 CFR Part 100 contains both requirements and guidance on how to satisfy the requirements. For example,Section IV, *Required Investigations,* of Appendix A, states that investigations are required for vibratory ground motion, surface faulting, and seismically induced floods and water waves. Appendix A then provides detailed guidance on what constitutes an acceptable investigation. A similar situation exists in Section V,

• seismic and Geologic Design Bases,* of Appendix A.

Geoscience assessments require considerable latitude in judgment. This latitude in judgment is needed because of limitations in data and the state-of-the-art of geologic and seismic analyses and because of the rapid evolution taking place in the geosciences in tenns of accumulating knowledge and in modifying concepts. This need appears to have been recognized when the existing regulation was.developed. The existing regulation states that it is based on limited geophysical and geological information and will be revised as necessary when 110re complete information becomes available.

However, having geoscience assessments detailed and cast in a regulation has created difficulty for applicants and the staff in terms of inhibiting the use of needed latitude in judgment. Also, it has inhibited flexibility in applying basic principles to new situations and the use of evolving methods of analyses (for instance, probabilistic) in the licensing process.

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The final regulation is streamlined, becoming a new section in Subpart B to 10 CFR Part 100 rather than a new appendix to Part 100. Also, the level of detail presented in the final regulation is reduced considerably. Thus, the final regulation contains: (a) required definitions, (b) a requirement to detennine the geological, seismological, and engineering characteristics of the proposed site, and (c) requirements to detennine the Safe Shutdown Earthquake Ground Motion (SSE), to determine the potential for surface defonnation, and to determine the design bases for seismically induced floods and water waves. The guidance documents describe how to carry out these required determinations. The key elements of the approach to detennine the SSE are presented in the following section. The elements are the guidance that is described in Regulatory Guide 1.165, "Identification and Characterization of Seismic Sources and Detennination of Safe Shutdown Earthquake Ground Motions.a

3. Uncertainties and Probabilistic Methods The existing approach for determining a Safe Shutdown Earthquake Ground Motion (SSE) for a nuclear reactor site, embodied in Appendix A to 10 CFR Part 100, relies on a *deterministic* approach. Using this deterministic approach, an applicant develops a single set of earthquake sources, develops for each source a postulated earthquake to be used as the source of ground motion that can affect the site, locates the postulated earthquake according to prescribed rules, and then calculates ground motions at the site.

Although this approach has worked reasonably well for the past two decades, in the sense that SSEs for plants sited with this approach are judged 31

to be suitably conservative, the approach has not explicitly recognized uncertainties in geosc1ences parameters. Because of uncertainties about earthquake phenomena {especially in the eastern United States}, there have often been differences of opinion and differing interpretations among experts as to the largest earthquakes to be considered and ground-motion models to be used, thus often making the licensing process relatively unstable.

Over the past decade, analysis methods for incorporating these different interpretations have been developed and used. These *probabilisticn methods have been designed to allow explicit incorporation of different models for zonation, earthquake size, ground motion, and other parameters. The advantage of using these probabilistic methods is their ability not only to incorporate different IROdels and different data sets, but also to weight them using judg-ments as to the validity of the different models and data sets, and thereby providing an explicit expression for the uncertainty in the ground motion estimates and a means of assessing sensitivity to various input parameters.

Another advantage of the probabilistic method is the target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants.

The final regulation explicitly recognizes that there are inherent uncertainties in establishing the seismic and geologic design parameters and allows for the option of using a probabilistic seismic hazard methodology capable of propagating uncertainties as a means to address these uncertainties. The rule further recognizes that the nature of uncertainty and the appropriate approach to account for it depend greatly on the tectonic regime and parameters, such as, the knowledge of seismic sources, the existence of historical and recorded data, and the understanding of tectonics.

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Therefore, methods other than the probabilistic methods, such as sensitivity analyses, may be adequate for some sites to account for uncertainties.

Methods acceptable to the NRC staff for implementing the regulation are described in Regulatory Guide 1.165, *Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion.*

The key elements of this approach are:

Conduct site-specific and regional geoscience investigations, Target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants, Conduct probabilistic seismic hazard analysis and determine ground motion level corresponding to the target exceedance probability Determine if information from the regional and site geoscience investigations change probabilistic results, Deteniine site-specific spectral shape and scale this shape to the ground motion level determined above, NRC staff review using all available data including insights*and information from previous licensing experience, and Update the data base and reassess probabilistic methods at least every ten years.

Thus, the approach requires thorough regional and site-specific geoscience investigations. Results of the regional and site-specific investigations must be considered in applications of the probabilistic method. The current probabilistic methods, the NRC sponsored study conducted by Lawrence Livermore National Laboratory (LLNL) or the Electric Power Research Institute (EPRI) 33

seismic hazard study, are regional studies without detailed information on any specific location. The regional and site-specific investigations provide detailed information to update the database of the hazard methodology as necessary.

It is also necessary to incorporate local site geological factors such as structural geology, stratigraphy, and topography and to account for site-specific geotechnical properties in establishing the design basis ground IROtion. In order to incorporate local site factors and advances in ground motion attenuation models, ground motion characteristics are determined using the procedures outlined in Standard Review Plan Section 2.5.2, Vibratory Ground Motion," Revision 3.

0 The NRC staff's review approach to evaluate ground motion estimates is described in SRP Section 2.5.2, Revision 3. This review takes into account the infonnation base developed in licensing more than 100 plants. Although the basic premise in establishing the target exceedance probability is that the current design levels are adequate, a staff review further assures that there is consistency with previous licensing decisions and that the scientific bases for decisions are clearly understood. This review approach will also assess the fairly complex regional probabilistic modeling, which incorporates multiple hypotheses and a multitude of parameters. Furthermore, the NRC staff's Safety Evaluation Report should provide a clear basis for the staff's decisions and facilitate communication with nonexperts ..

4. Safe Shutdown Earthquake.

34

The existing regulation (10 CFR Part 100, Appendix A, Section V(a)(l)(iv)) states *The maximum vibratory accelerations of the Safe Shutdown Earthquake at each of the various foundation locations of the nuclear power plant structures at a given site shall be detemined .** n The location of the seismic input 110tion control point as stated in the existing regulation has led to confrontations with many applicants that believe this stipulation is inconsistent with good engineering fundamentals.

The final regulation RlOVes the location of the seismic input motion control point from the foundation-level to the free-field at the free ground surface. The 1975 version of the Standard Review Plan placed the control mot..:!1>n in the free-field. The final regulation is also consistent with the resolution of Unresolved Safety Issue (USI) A-40, *seismic Design Criteri.a" (August 1989), that resulted in the revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3. The final regulation also requires that the horizontal component of the Safe Shutdown Earthquake Ground Motion in the free-field at the foundation level of the structures must be an appropriate response spectrum considering the site geotechnical properties, with a peak ground acceleration of at least O.lg.

s. Value of the Operating Basis Earthquake Ground Motion (QBE) and Required QBE Analyses.

The existing regulation (10 CFR Part 100, Appendix A, Section V(a){2))

states that the maximum vibratory ground B10tion of the OBE is at least one half the maximum vibratory ground motion of the Safe Shutdown Earthquake ground motion. Also, the existing regulation (10 CFR Part 100, Appendix A, 35

Section Vl{a){2)) states that the engineering method used to insure that structures, systems, and co11tponents are capable of withstanding the effects of the OBE shall involve_ the use of either a suitable dynamic analysis or a suitable qualification test. In some cases, for instance piping, these multi-facets of the OBE in tne existing regulation made it possible for the OBE to have more design significance than the SSE. A decoupling of the OBE and SSE has been suggested in several documents. For instance, the NRC staff, SECY-79-300, suggested that a compromise is required between design for a broad spectrum of unlikely events and optimum design for normal operation.

Design for a single limiting event {the SSE) and inspection and evaluation for earthquakes in excess of some specified limit {the OBE), when and if they occur, may be the most sound regulatory approach. NUREG-1061, "Report of the U.S. Nuclear Regulatory Comission Piping Review Connittee," Vol.5, April 1985, {Table IO.I) ranked a decoupling of the OBE and SSE as third out of six high priority changes. In SECY-90-016, *Evolutionary Light Water Reactor

• (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements," the NRC staff states that it agrees that the QijE should not control the design of safety systems. Furthermore, the final safety evaluation reports related to the certification of the System 80+ and the Advanced Boiling Water Reactor design {NUREG-1462 and NUREG-1503, respectively) have already adopted the single earthquake design philosophy.

Activities equivalent to OBE-SSE decoupling are also being done in foreign countries. For instance, in Germany their new design standard requires only one design basis earthquake {equivalent to the SSE). They require an inspection-level earthquake {for shutdown) of 0.4 SSE. This level 36

was set so that the vibratory ground 1ROtion should not induce stresses exceeding the allowable stress limits originally required for the QBE design.

The final regulation allows the value of the QBE to be set at (i) ane-third or less of the SSE~ where OBE requirements are satisfied without an explicit response or design analyses being perfonned, or (ii) a value greater than one-third of the SSE, where analysis and design are required. There are two issues the applicant should consider in selecting the value of the OBE:

first, plant shutdown is required if vibratory ground motion exceeding that of the OBE occurs (discussed below in Item 6, Required Plant Shutdown), and second, the amount of analyses associated with the OBE. An applicant may detennine that at one-third of the SSE level, the probability of exceeding the QBE vibratory ground inotion is too high, and the cost associated with plant shutdown for inspections and testing of equipment and structures *prior to restarting the plant is unacceptable. Therefore, the applicant may voluntarily select an OBE value at some higher fraction of the SSE to avoid plant shutdowns. However, if an applicant selects an OBE value at*a fraction of the SSE higher than one-third, a suitable analysis shall be perfonned to demonstrate that the requirements associated with the OBE are satisfied. The design shall take into account soil-structure interaction effects and the expected duration of the vibratory ground motion. The requirement associated with the OBE is that all structures, systems, and components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public shall remain functional and within applicable stress, strain and deformation limits when subjected to the effects of the OBE in combination with normal operating loads.

37

As stated, it is detennined that if an OBE of one-third or less of the SSE. is used, the requirements of the OBE can be satisfied without the applicant perfoming any explicit response analyses. In this case, the OBE serves the function of an inspection and shutdown earthquake. Some minimal design checks and the applicability of this position to seismic base isolation of buildings are discussed below. There is high confidence that, at this ground-110tion level with other postulated concurrent loads, most critical structures, systems, and components will not exceed currently used design.

limits. This is ensured, in part, because PRA insights will be used to support a aargins-type assessment of seismic events. A PRA-based seismic margins analysis will consider sequence-level High Confidence, Low Probability of Failures (HCLPFs) and fragilities for all sequences leading to core damage or contain111ent failures up to approximately one and two-thirds the ground motion acceleration of the design basis SSE (

Reference:

Item II.N, Site-Specific Probabilistic Risk Assessment and Analy$1s of External Events, 111e1110randua fr011 Samuel J. Chilk to James M. Taylor,

Subject:

SECY-93-087 -

Policy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR) Designs, dated July 21, 1993).

There are situations associated with current analyses where only the OBE is associated with the design requirements, for example, the ultimate heat sink (see Regulatory Guide 1.27, *ultimate Heat Sink for Nuclear Power Plants*). In these situations, a value expressed as a fraction of the SSE response would be used in the analyses.Section VII of this final rule identifies existing guides that would be revised technically to maintain the existing design philosophy.

38

In SECY-93-087, HPolicy, Technical, and Licensing Issues Pertaining to Evolutionary and Advance Light-Water Reactor (ALWR) Designs,~ the NRC staff requested COfllllission approval on 42 technical and policy issues pertaining to either evolutionary LWRs, passive LWRs, or both. The issue pertaining to the elimination of the OBE is designated I.M. The NRC staff identified actions necessary for the design of structures, systems, and colllJ)onents when the OBE design requirement is eliminated. The NRC staff clarified that guidelines should be maintained to ensure the functionality of components, equipment, and their supports. In addition, the NRC staff clarified how certain design requirements are to be considered for buildings and structures that are currently designed for the DBE, but not the SSE. Also, the NRC staff has evaluated the effect on safety of eliminating the OBE from the design load cOllbinations for selected structures, systems, and components anw has developed proposed criteria for an analysis using only the SSE. Commission approval is documented in the Chilk to Taylor memorandum dated July 21, 1993, cited above.

More than one earthquake response analysis for a seismic base isolated nuclear power plant design may be necessary to ensure adequate performance at all earthquake levels. Decisions pertaining to the response analyses associated with base isolated facilities will be handled on a case by case basis.

6. Regu1red Plant Shutdown.

The current regulation (Section V(a){2)) states that if vibratory ground motion exceeding that of the OBE occurs, shutdown of the nuclear power plant 39

will be required. The supplementary information to the final regulation (published November 13, 1973; 38 FR 31279, Item 6e} includes the following statement: *A footnote has been added to sS0.36(c)(2} of 10 CFR Part 50 to assure that each power plant is aware of the limiting condition of operation which is imposed under Section V(2) of Appendix A to 10 CFR Part 100. This limitation requires that if vibratory ground motion exceeding that of the OBE occurs, shutdown of the nuclear power plant will be required. Prior to resuming operations, the licensee will be required to demonstrate to the Comission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public." At that time, it was the intention of the Commission to treat the OBE as a limiting condition of operation. From the statement in the Supplementary Information, the Comission directed applicants to specifically review 10 CFR Part 100 to be aware of this intention in complying with the requirements of 10 CFR 50.36. Thus, the requirement to shut down if an OBE occurs was expected to be implemented by being included among the technical specifications submitted by applicants after the adoption of Appendix A. In fact, applicants did not include OBE shutdown requirements in their technical specifications.

The final regulation treats plant shutdown associated with vibratory ground motion exceeding the OBE or significant plant damage as a condition in every operating license. A new s50.54(ff) is added to the regulations to require a process leading to plant shutdown for licensees of nuclear power plants that COIIIPlY with the earthquake engineering criteria in Paragraph IV(a)(3} of Appendix S, *Earthquake Engineering Criteria for Nuclear Power Plants,* to 10 CFR Part 50. ID111ediate shutdown could be required until it is 40 r

determined that structures, systems, and components needed for safe shutdown are still functional.

Regulatory Guide 1.166, "Pre-Earthquake Planning and Inmediate Nuclear Power Plant Operator Post-Earthquake Actions,* provides guidance acceptable to the NRC staff for deten11i~ir.~ whether or not vibratory ground motion exceeding the OBE ground motion or significant plant damage had occurred and the timing of nuclear power plant shutdown. The guidance is based on criteria developed by the Electric.Power Research Institute (EPRI). The decision to shut down the plant should be made by the licensee within eight hours after the earthquake. The data from the seismic instrumentation, coupled with information obtained from a plant walk down, are used to make the determina-tion of when the plant should be shut down, if it has not already been shut down by*operational perturbations resulting from the seismic event. The guidance in Regulatory Guide 1.166 is based on two assumptions, first, that the nuclear power plant has operable seismic instrumentation, including the equipment and software required to process the data within four hours.after an earthquake, and second, that the operator walk down inspections can be .

performed in approximately four to eight hours depending on the number of personnel conducting the inspection. The regulation also includes a provision that requires the licensee to consult with the C0tm1ission *and to propose a plan for the timely, safe shutdown of the nuclear power plant if systems, structures, or components necessary for a safe shutdown or to maintain a safe shutdown are not available.

Regulatory Guide 1.167, "Restart of a Nuclear Power Plant Shut Down by a Seismic Event,* provides guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and 41

structures prior to plant restart. This guidance is also based on EPRI reports. Prior to resuming operations, the licensee must demonstrate to the COllll'lission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public. The results of post-shutdown inspections, operability checks, and surveillance tests must be docU11ented in written reports and submitted to the Director, Office of Nuclear Reactor Regulation. The licensee shall not resume operation untfl authorized to do so by the Director, Office of Nuclear Reactor Regulation.

7. Clarify interpretations.

Section 100.23 resolves questions of interpretation. As an example, definitions and required investigations stated in the final regulation do not contain the phrases in Appendix A to Part 100 that were more applicable to only the western part of the United States.

The institutional definition for ~safety-related structures, systems, and components" is drawn from Appendix A to Part 100 under III(c) and Vl(a).

With the relocation of the earthquake engineering criteria to Appendix S to Part 50 and the relocation and modification to dose guidelines in s50.34(a}(l}, the definition of safety-related structures, systems, and

• components is included in Part 50 definitions with references to both the Part 100 and Part 50 dose guidelines.

YI. Related Regulatory Guides and Standard Review Plan Sections 42

The NRC is developing the following regulatory guides and standard review plan sections to provide prospective licensees with the necessary guidance for illf)leaenting the final regulation. The notice of availability for these materials will be published in a later issue of the Federal Register.

1. Regulatory Guide 1.165, *Ident1f1cation and Characterization of Seis1c Sources and Determination of Shutdown Earthquake Ground Motions.* The guide provides general guidance and rec011111endations, describes acceptable procedures and provides a list of references that present acceptable methodologies to identify and characterize capable tectonic sources and seismogen1c sources.Section V.B.3 of this rule describes the key elements.

2. Regulatory Guide 1.12, Revision 2, *Nuclear Power Plant lnstruaentation for Earthquakes.* The guide describes seismic instrumentation type and location, operability, characteristics, installation, actuation, and maintenance that are acceptable to the NRC staff.

3. Regulatory Guide 1.166, *Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions.* The guide provides guidelines that are acceptable to the NRC staff for a timely evaluation of the recorded seismic instrumentation data and to determine whether or not plant shutdown is required.

4. Regulatory Guide l.~67, *Restart of a Nuclear Power Plant Shut Down by a Seismic Event.* The guide provides guidelines that are acceptable to the NRC staff for perfonaing inspections and tests of nuclear power plant equipment and structures prior ~o restart of a plant that has been shut down because of a seismic event.

43

5. Standard Review Plan Section 2.5.i, Revision 3, "Basic Geologic ar.1 Seismic Information.* This SRP Section describes procedures to assess the adequacy of the geologic and seismic information cited in support of the applicant's conclusions concerning the suitability of the plant site.

6. Standard Review Plan Section 2.5.2, Revision 3 "Vibratory Ground Motion.* This SRP Section describes procedures to assess the ground motion potential of seismic sources at the site and to assess the adequacy of the SSE.

7. Standard Review Plan Section 2.5.3, Revision 3, *surface Faulting."

This SRP Section describes procedures to assess the adequacy of the applicant's submittal related to the existence of a potential for surface faulting affecting the site.

8. Regulatory Guide 4.7, Revision 2, "General Site Suitability Criteria for Nuclear Power Plants." This guide discusses the major site characteristics related to public health and safety and environmental issues that the NRC staff considers in determining the suitability of sites.

VII. Future Regulatory Action Several existing regulatory guides will be revised to incorporate editorial changes or maintain the existing design or analysis philosophy.

These guides w111 be issued as final guides without public coment subsequent to the publication of the final regulations.

The following regulatory guides will be revised to incorporate editorial changes, for example to reference new sections to Part 100 or Appendix S to Part 50. No technical changes will be made in these regulatory guides.

44

1. 1.57, "Design Limits and Loading Combinations for Metal Primary Reactor Containment System Collll)onents."

2. 1.59, *oesign Basis Floods for Nuclear Power Plants."

3. 1.60, *Design Response Spectra for Seismic Design of Nuclear Power Plants."

4. 1.83, *Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes.*

5. 1.92, *combining Modal Responses and Spatial Components in Seismic Response Analysis."

6. 1.102, *flood Protection for Nuclear Power Plants."

7. 1.121, *eases for Plugging Degraded PWR Steam Generator Tubes."

8. 1.122, "Development of Floor Design Response Spectra for Sei.smic Design of Floor-Supported Equipment or Components.*

The following regulatory guides will be revised to update the design or analysis philosophy, for example, to change QBE to a fraction of the SSE:

1. 1.3, *Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors.*

2. 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors.*

3. 1.27, *ultimate Heat Sink for Nuclear Power Plants.*

45

4. 1.100, "Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants."

5. 1.124, *service Limits and Loading Combinations for Class 1 Linear-Type Component Supports."

6. 1.130, *service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports.*

7. 1.132, *site Investigations for Foundations of Nuclear Power Plants."

8. 1.138, *Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants."

9. 1.142, *safety-Related Concrete Structures for Nuclear Power C

Plants (Other than Reactor Vessels and Containments).*

10. 1.143, "Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-WateHooled Nuclear Power Plants."

Minor and confoming changes to other Regulatory Guides and standard review plan sections as a result of changes in the nonseismic criteria are also planned. If substantive changes are made during the revisions, the applicable guides will be issued for public co11111ent as draft guides.

VIII. Referenced Documents An interested person may examine or obtain copies of the documents referenced in this rule as set out below.

46

Copies of NUREG-0625, NUREG-1061, NUREG-1150, NUREG-1451, NUREG-1462, NUREG-1503, and NUREG/CR-2239 may be purchased fr011 the Superintendent of Doc1.111ents, U.S. Governaent Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. Copies also are available from the National Technical Infomat1on Service, 5285 Port Royal Road, Springfield, VA 22161. A copy also is available for inspection and copying for a fee in the NRC Public Document Room, 2120 L Street, NW. (Lower Level), Washington, DC.

Copies of issued regulatory guides may be purchased fr0111 the Government Printing Office {GPO) at the current GPO price. Infomation on current GPO prices may be obtained by contacting the Superintendent of Documents, U.S.

Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328.

Issued guides also may be purchased from the National Technical Infonnati"on Service on a standing order basis. Details on this service may be obtained by writing NTIS, 5826 Port Royal Road, Springfield, VA 22161.

SECY 79-300, SECY 90-016, SECY 93-087, and WASH-1400 are available for inspection and copying for a fee at the NRC Public Document Room, 2120 L Street, NW. {Lower Level), Washington, DC.

IX. Suaary of Conaents on the Proposed Regulations.

A. Reactor Siting Criteria (Nonseisaic).

47

Eight organizations or individuals comented on the nonseismic aspects of the second proposed revision. The first proposed revision issued for connent in October 20, 1992, (57 FR 47802) elicited strong comments in regard to proposed numerical values of population density and a minimum distance to the exclusion area boundary (EAB) in the rule. The second proposed revision (October 17, 1994; 59 FR 52255) would delete these from th~ rule by providing guidance on population density in a Regulatory Guide and determining the distance to the EAB and LPZ by use of source tenn and dose calculations. The rule would contain basic site criteria, without any numerical values.

Several c011111entors representing the nuclear industry and international nuclear organizations stated that the second proposed revision was a significant improvement over the first proposed revision, while the only public interest group connented that the NRC had retreated from decoupling siting and design in response to the comments of foreign entities.

Most comnents on the second proposed revision centered on the use of total effective dose equivalent (TEDE), the proposed single numerical dose acceptance criterion of 25 rem TEDE, the evaluation of the maximum dose in any two-hour P,,eriod, and the question of whether an organ capping dose should be adopted.

Virtually all con111enters supported the concept of TEDE and its use.

However, there were differing views on the proposed numerical dose of 25 rem and the proposed use of the maximum two-hour period to evaluate the dose.

Virtually all industry connenters felt that the proposed numerical value of 25 rem TEDE was too low and that it represented a "ratchet" since the use of the current dose criteria plus organ weighting factors would suggest a value of 34 rem TEDE. In addition, all industry commenters believed the "sliding" two-hour 48

window for dose evaluation to be confusing, illogical and inappropriate. They favored a rule that was based upon a two hour period after the onset of fission product release, similar in concept to the existing rule. All industry cementers opposed the use of an organ capping dose. The only public interest group that co111PPntP.d did not object to the use of TEDE, favored the proposed dose value of 25 rea, and supported an organ capping dose.

B. Seis*ic and Earthquake Engineering Criteria.

Seven letters were received addressing either the regulations or both the regulations and the draft guidance docU11ents identified in Section VI

{except DG-4003). An additional five letters were received addressing only the guidance documents, for a total of twelve co11111ent letters. A document,

• Resolution of Public Connents on the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," is available explaining the NRC's disposition of the connents received on the regulations. A copy of this document has been placed in the NRC Public Document Room, 2120 L Street NW.

{Lower Level), Washington, DC. Single copies are available from Dr. Andrew J.

Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Connission, Washington, DC 20555-0001, telephone (301) 415-6010. A second docuaent, "Resolution of Public Comnents on Draft Regulatory Guides and Standard Review Plan Sections Pertaining to the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," will explain the NRC's disposition of the coanents received on the guidance documents. The Federal Register notice announcing the avaliability of the guidance documents will also discuss how to obtain copies of the corrment resolution document.

49

A sWllllary of the major connents on the proposed regulations follows.

Supplementary Information Section III, Genesis (Application)

Coatent: The Departllent of Energy (Office pf Civilian Radioactive Waste Manage11ent), requests an explicit statement on whether or nots 100.23 applies to the Mined Geologic Disposal System (MGDS) and a Monitored Retrievable Storage (MRS) facility. The NRC has noted in NUREG-1451, "Staff Technical Position on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Respository,* that Appendix A to 10 CFR Part 100 does not apply to a geologic repository. NUREG-1451 also notes that the contemplated revisions to Part 100 wo11ld also not be applicable to a geologic repository. Section 72.102(b) requires that, for an MRS located west of the Rocky Mountain front or in areas of known potential seismic activity in the east, the seismicity be evaluated by the techniques of Appendix A to 10 CFR Part 100.

Response: Although Appendix A to 10 CFR Part 100 is titled *seismic and Geologic Siting Criteria for Nuclear Power Plants,* it is also referenced in two other parts of the regulation. They are (1) Part 40, "Domestic Licensing of Source Material,* Appendix A, *criteria Relating to the Operation of Uranium Mills and the Disposition of Tailings or Waste Produced by the Extraction or Concentration of Source Material from Ores Processed Primarily for Their Source Material Content,* Section I, Criterion 4{e), and (2) Part 72, *Licensing Requirements for the Independent Storage of Spent Nuclear Fuel 50

and High-Level Radioactive Waste,* Paragraphs {a){2), {b) and {f){l) of s72.102.

The referenced applicability of s 100.23 to other than power reactors, if considered appropriate by the NRC, would be a separate rulemaking. That rulemaking would clearly state the applicability of s 100.23 to an MRS or other facility. In addition, NUREG-1451 will remain the NRC staff technical position on seismic siting issues pertaining to an MGDS until it is superseded through a rulemaking, revision of NUREG-1451, or other appropriate mechanism.

Section V{B){S), "Value of the Operating Basis Earthquake Ground Motion {OBE) and Required OBE Analysis."

Coment: One comenter, ABB Combustion Engineering Nuclear Systems, specifically stated that they agree with the NRC's proposal to not require explicit design analysis of the OBE if its peak acceleration is less than one-third of the Safe Shutdown Earthquake Ground Motion {SSE). The only negative c01111ents, fr0111 6.C. Slagis Associates, stated that the proposed rule in the area of required OBE analysis is not sound, not technically justified, and not appropriate for the design of pressure-retaining components. The following are specific connents {limited to the design of pressure-retaining components to the ASHE Boiler and Pressure Vessel Section Ill rules) that pertain to the suppleaental information to the proposed regulations, item V{B){S), "Value of the Operating Basis Earthquake Ground Motion {OBE) and Required OBE Analysis."

{l) Connent: Disagrees wi+h the statement in SECY-79-300 that design for a single limiting event and inspection and evaluation for earthquakes in excess of s011e specified limit may be the 110st sound regulatory approach. It 51

is not feasible to inspect for cyclic damage to all the pressure-retaining components. Visually inspecting for pemanent deformation, or leakage, or failed component supports is certainly not adequate to determine cyclic damage.

Response: The NRC agrees. Postearthquake inspection and evaluation guidance is described in Regulatory Guide 1.167 (Draft was D6-1035), *Restart of a Nuclear Power Plant Shut Down by an Seismic Event.* The guidance is not liaited to visual inspections; it includes inspections, tests, and analyses including fatigue analysis. *

(2) Coament: Disagrees with the NRC statement in SECY-090-016 that the OBE should not control design. There is a problem with the present require11ents. Requiring design for five OBE events at one-half SSE is unrea~istic for aost {all?) sites and requires an excessive and unnecessary number of seismic supports. The solution is to properly define the OBE magnitude and the nulllber of events expected during the life of the plant and to require design for that loading. OBE may or may not control the design.

But you cannot asswae, before you have the seismicity defined and before you have a component design, that OBE will not govern the design.

Response: The NRC has concluded that design requirements based on an estimated OBE magnitude at the plant site and the number of events expected during the plant life will lead to low design values that will not control the design, thus resulting in unnecessary analyses.

(3) Coanent: It is not technically justified to assume that Section III components will reaain within applicable stress limits (Level B limits) at one-third the SSE. The Section III acceptance criteria for Level D {for an SSE) is completely different than that for Level B {for an OBE). The Level D 52

criteria is based on surviving the extremely-low probability SSE load. Gross structural deformations are possible, and it is expected that the component will have to be replaced. Cyclic effects are not considered. The cyclic effects of the repeated earthquakes have to be considered in the design of the c011ponent to ensure pressure boundary integrity throughout the life of the component, especially if the SSE can occur after the lower level earthquakes.

Response: In SECY-93-087, Issue I.M, *Elimination of Operating-Basis Earthquake,* the NRC recognizes that a designer of piping systems considers

• the effects of primary and secondary stresses and evaluates fatigue caused by repeated cycles of loading. Primary stresses are induced by the inertial effects of vibratory motion. The relative motion of anchor points induces secondary stresses. The repeating seismic stress cycles induce cyclic effects (fatigue). However, after reviewing these aspects, the NRC concludes that, for primary stresses, if the OBE is established at one-third the SSE, the SSE load combinations control the piping design when the earthquake contribution d011inates the load collbination. Therefore, the NRC concludes that eliminating the OBE piping stress load combination for primary stresses in piping systems will not significantly reduce existing safety margins.

Eliminating the OBE will, however, directly affect the current methods used to evaluate the adequacy of cyclic and secondary stress effects in the piping design. Eliminating the OBE from the load combination could cause uncertainty in evaluating the cyclic (fatigue) effects of earthquake-induced motions in piping systems and the relative motion effects of piping anchored to equipment and structures at various elevations because both of these effects are currently evaluated only for OBE loadings. Accordingly, to account for earthquake cycles in the fatigue analysis of piping systems, the 53

staff proposes to develop guidelines for selecting a number of SSE cycles at a fraction of the peak a11plitude of the SSE. These guidelines will provide a level of fatigue design for the piping equivalent to that currently provided in Standard Review Plan Section 3.9.2.

Positions pertaining to the elimination of the DBE were proposed in SECY-93-087. C011111ission approval is docuented in a memorandum from Samuel J.

Chilk to James M. Taylor,

Subject:

SECY-93-087 - Policy, Technical and Licensing Issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs, dated July 21, 1993.

(4) Coanent: There is one major flaw in the "SSE only" design approach.

The equipm!nt designed for SSE is limited to the equipment necessary to assure the integrity of the reactor coolant pressure boundary, to shutdown the reactor, and to prevent or mitigate accident consequences. The equipment designed for SSE is only part of the equipment anecessary for continued operation without undue risk to the health and safety of the public.* Hence, by this rule, it is possible that some equipment necessary for continued operation will not be designed for SSE or DBE effects.

Response: The NRC does not agree that the design approach is flawed. It is not possible that some equipment necessary for continued llfi operation will not be designed for SSE or OBE effects. General Design Criterion 2, "Design Bases for Protection Against Natural Phenocnena," of Appendix A,

• General Design Criteria for Nuclear Power Plants,* to 10 CFR Part 50 requires that nuclear power plant structures, systems, and components important to safety be designed to withstand the effects of earthquakes without loss of capability to perform their safety functions. The criteria in Appendix S to

\

10 CFR Part 50 implement General Design Criterion 2 insofar as it requires 54

structures, systems, and coaponents important to safety to withstand the effects of earthquakes. Regulatory Guide 1.29, *seismic Design Classification,* describes a method acceptable to the NRC for identifying and classifying those features of light-water-cooled nuclear power plants that should be designed to withstand the effects of the SSE. Currently, components which are designed for OBE only include components such as waste holdup tanks.

As noted in Section VII, Future Regulatory Actions, regulatory guides related to these components will be revised to provide alternative design requirements.

10 CFR 100.23 The Nuclear Energy Institute (NEI) congratulated the NRC staff for carefully considering and responding to the voluminous and complex co11111ents that were provided on the earlier proposed rulemaking package (October 20, 1992; 57 FR 47802) and considered that the seismic portion of the proposed rule11aking package is nearing maturity and with the inclusion of industry's cOlllll8nts (which were principally on the guidance documents), has the potential to satisfy the objectives of predictable licensing and stable regulations.

Both NEI and Westinghouse Electric Corporation support the regulation format, that is, prescriptive guidance is located in regulatory guides or standard review plan sections and not the regulation.

NEI and Westinghouse Electric Corporation support the removal of the requirement from the first proposed rulemaking (57 FR 47802) that both 55

detenninistic and probabilistic evaluations must be conducted to detennine site suitability and seismic design requireaents for the site. [Note: the c01111enters do not agree with the NRC staff's detenninistic check of the seismic sources and parameters used in.the LLNL and EPRI probabilistic seismic hazard analyses (Regulatory Guide 1.165, draft was DG-1032). Also, they do not support the NRC staff's detenninistic check of the applicants subllittal (SRP Section 2.5.2). These items are addressed in the document pertaining to comaent resolution of the draft regulatory guides and standard review plan sections.]

Conment: NEI, Westinghouse Electric Corporation, and Yankee Atomic Electric Corporation rec011111end that the regulation should state that for existing sites east of the Rocky Mountain Front (east of approximately 105° west longitude), a 0.3g standardized design level is acceptable at these sites given confirmatory foundations evaluations [Regulatory Guide 1.132, but not the geologic, geophysical, seismological investigations in Regulatory Guide 1.165].

Response: The NRC has detennined that the use of a spectral shape anchored to 0.3g peak ground acceleration as a standardized design level would be appropriate for existing central and eastern U.S. sites based on the current state of knowledge. However, as new infonnation becomes available it may not be appropriate for future licensing decisions. Pertinent information such as that described in Regulatory Guide 1.165 (Draft was DG-1032) is needed to make that assessment. Therefore, it is not appropriate to codify the request.

56

COllll8nt: NEI rec0111118nded a rewording of Paragraph (a), Applicability.

Although unlikely, an applicant for an operating license already holding a construction pennit lllaY elect to apply the amended methodology and criteria in Subpart B to Part 100.

Response: The NRC ~111 address this request on a case-by-case basis rather than through a generic change to the regulations. This situation pertains to a limited number of facilities in various stages of construction.

Some of the issues that 111.1st be addressed by the applicant and NRC during the operating license review include differences between the design bases derived fr011 the current and amended regulations (Appendix A to Part 100 ands 100.23, respectively), and earthquake engineering criteria such as, OBE design requirements and OBE shutdown requirements.

Appendix S to 10 CFR Part 50 Support for the NRC position pertaining to the elimination of the Operating Basis Earthquake Ground Motion (OBE) response analyses has been documented in various NRC publications such as SECV-79-300, SECV-90-016, *SECV-93-087, and NUREG-1061. The final safety evaluation reports related to the certification of the Syste111 80+ and the Advanced Boiling Water Reactor design (NUREG-1462 and NUREG-1503, respectively) have already adopted the single earthquake design philosophy. In addition, similar activities are being done in foreign countries, for instance, Gennany. (Additional discussion is provided in Section V(B)(5) of this rule).

57

Coaent: The American Society of Civil Engineers (ASCE} reconmended that the seismic design and engineering criteria of ASCE Standard 4, "Seismic Analysis of Safety-Related Nuclear Structures and Coanentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures," be incorporated by reference into Appendix s to 10 CFR Part 50.

Response: The Co1111ission has determined that new regulations will be more stre1111lined and contain only basic requirements with guidance being provided in regulatory guides and, to some extent, in standard review plan sections. Both the NRC and industry have experienced difficulties in applying prescriptive regulations such as Appendix A to 10 CFR Part 100 because they inhibit the use of needed latitude in judgement. Therefore, it is c0111110n NRC practice not to reference publications such as ASCE Standard 4 (an analysis, not design standard} in its regulations. Rather, publications such as ASCE Standard 4 are cited 1n regulatory guides and standard review plan sections.

ASCE Standard 4 is cited in the 1989 revision of Standard Review Plan Sections 3.7.1, 3.7.2, and 3.7.3.

COlllll8nt: The Department of Energy stated that the required consideration of aftershocks in Paragraph IY(B), Surface Deformation, is confusing and reconaended that it be deleted.

Response: The NRC agrees. The reference to aftershocks in Paragraph IV(b) has been deleted. Paragraphs VI(a), Safe Shutdown Earthquake, and Vl(B}(3) of Appendix A to Part 100 contain the phrase "including aftershocks."

The *including aftershocks* phrase was removed from the Safe Shutdown Earthquake Ground Motion requirements in the proposed regulation. The recomended change will make Paragraphs IV(a)(l}, *safe Shutdown Earthquake 58

Ground Motion,* and IV(b), *surface Defonnation, of Appendix S to 10 CFR Part 50 consistent.

X. Saall Business Regulatory Enforceaent Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness Act of 1996 the NRC has determined that this action is not a major rule and has verified this detenaination with the Office of Information and Regulatory Affairs of 0MB.

XI. Finding of No Significant Env1rof1118ntal Impact: Availability The C0n111ission has determined under the National Environmental Policy Act of 1969, as amended, and the Co11111ission's regulations in Subpart A of 10 CFR Part 51, that this regulation is not a major Federal action significantly affecting the quality of the human environment and therefore an environmental impact statement is not required.

The revisions associated with the reactor siting criteria in 10 CFR Part 100 and the relocation of the plant design requirements from 10 CFR Part 100 to 10 CFR Part 50 have been evaluated against the current requirements. The C011111ission has concluded that relocating the requirement for a dose calculation to Part 50 and adding more specific site criteria to Part 100 does not decrease the protection of public health and safety over the current regulations. The amendments do not affect nonradiological plant effluents and have no other environmental impact.

59

The addition of sl00.23 to 10 CFR Part 100, and the addition of Appe~1ix S to 10 CFR Part 50, will not change the radiological environmental impact offsite. Onsite occupational radiation exposure associated with inspection and maintenance will not change. These activities are principally associated with base line inspections of structures, equipment, and piping, and with mintenance of seisic instrU11entation. Baseline inspections are needed to differentiate between pre-existing conditions at the nuclear power plant and earthquake related damage. The structures, equipment and piping selected for these inspections are those routinely examined by plant operators during normal plant walkdowns and inspections. Routine maintenance of seismic instrU111entation ensures its operability during earthquakes. The location of the seismic instrumentation is similar to that in the existing nuclear power plants. The U1endments do not affect nonradiological plant effluents and have no other environmental impact.

The environmental assessment and finding of no significant impact on which this determination is based are available for inspection at the NRC Public Docwitent Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environlll8ntal assessment and finding of no significant impact are available from Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory COflll1ission, Washington, DC 20555-0001, telephone (301) 415-6010.

XII. Paperwork Reduction Act Statement This final rule amends information collection requirements that are subject to the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

60

These requir81118nts were a~~roved by the Office of Management and Budget, approval numbers 3150-0011 and 3150-0093.

The public reporting burden for this collection of information is estimated to average 800,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> per response, including the tiine for reviewing instructions, searching existing data sources, gathering and aaintaining the data needed, and c0111pleting and reviewing the collection of infonaation. Send c011118nts on any aspect of this collection of information, including suggestions for reducing the burden, to the Information and Records Management Branch (T-6 F33), U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, or by Internet electronic mail to BJSl@NRC.GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NEOB-10202, (3150-0011 and 3150-0093), Office of Management and Budget, Washington, DC 20503.

Public Protection Notification The NRC ay not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid 0MB control number.

XIII. Regulatory Analysis The Connission has prepared a regulatory analysis on this regulation.

The analysis exa11ines the costs and benefits of the alternatives considered by the Connission. Interested persons r.1y examine a copy of the regulatory analysis at the NRC Public Document Room, 2120 L Street NW. (Lower Level),

Washington, DC. Single copies of the analysis are available from Dr. Andrew 61

J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Connission, Washington, DC 20555-0001, telephone (301) 415-6010.

XIV. Regulatory Flex1b111ty Cert1f1cat1on As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b),

the C011111ission certifies that this regulation does not have a significant econ011ic illi)act on a substantial number of s all entities. This regulation affects only the licensing and operation of nuclear power plants. The companies that own these plants do not fall within the definition of nsmall entities* set forth in the Regulatory Flexibility Act or the size standards established by the NRC (April 11, 1995; 60 FR 18344).

XV. Backf1t Analysis The NRC has determined that the backfit rule, 10 CFR 50.109, does not apply to this regulation, and therefore, a backfit analysis is not required for this regulation because these amendments do not involve any provisions -

that would impose backfits as defined in 10 CFR 50.109(a)(l). The regulation would apply only to applicants for future nuclear power plant construction permits, preliminary design approval, final design approval, manufacturing licenses, early site reviews, operating licenses, and combined operating licenses.

List of SUbjects 62

10 CFR Part 21 - Nuclear power plants and reactors, Penalties, Radiation protection, Reporting and recordkeeping requirements.

10 CFR Part SO-Antitrust, Classified 1nfonnat1on, Criminal penalties, Fire protection, lntergovernaental relations, Nuclear power plants and reactors, Radiation protection, Reactor siting criteria, Reporting and recordkeeping requirements.

10 CFR Part 52 -Administrative practice and procedure, Antitrust, Backfitting, Combined license, Early site permit, Emergency planning, Fees, Inspection, Li ited work authorization, Nuclear power plants and reactors, Probabilistic risk assessment, Prototype, Reactor siting criteria, Redress of site, Reporting and recordkeeping requirements, Standard design, Standard design certification.

10 CFR Part 54 - Administrative practice and procedure, Age-related degradation, Backfitting, Classified information, Criminal penalties, EnvironRtental, Nuclear power plants and reactors, Reporting and recordkeeping requirements.

10 CFR Part 100 - Nuclear power plants and reactors, Reactor siting criteria.

For the reasons set out in the preamble and under the authority of the Atomic Energy Act of 1954, as amended, the Energy Reorganization Act of 1974, 63

as aaended, and 5 U.S.C. 552 and 553, the NRC is adopting the following 111endments to 10 CFR Parts 21, SO, 52, 54, and 100.

PART 21 - REPORTING OF DEFECTS Ari> NONCOIIPLIANCE

1. The authority citation for Part 21 continues to read as follows:

AUTHORITY: Sec. 161, 68 Stat. 948, as amended, sec. 234, 83 Stat. 444, as aaended, sec. 1701, 106 Stat. 2951, 2953 (42 U.S.C. 2201, 2282, 2297f);

secs. 201, as aaended, 206, 88 Stat. 1242, as aaended, 1246 (42 U.S.C. 5841, 5846).

Section 21.2 also issued under secs. 135, 141, Pub. L.97-425, 96 Stat.

2232, 2241 (42 u.s.c. 10155, 10161).

2. In s21.3, the definition for Basic COJRPooent (1)(1)(C) is revised to read as follows:

1 21.3 Deftn1t1ons.

Baste C01J10aent. (l)(i) * * *

(C) The capability to prevent or itigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to 1n sS0.34(a)(l) or sl00.11 of this chapter, as applicable.

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PART 50 - DOIIESTIC LICENSING OF PRODUCTION AtlJ UTILIZATION FACILITIES

3. The authority citation for Part SO continues to read as follows:

AUTHORITY: Secs. 102, 103, 104, 105, 161, 182, 183, 186, 189, 68 Stat.

936, 937, 938, 948, 953, 954, 955, 956, as amended, sec. 234, 83 Stat. 1244, as 1111ended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, as amended, 202, 206, 88 Stat. 1242, as amended, 1244, 1246, (42 u.s.c. 5841, 5842, 5846).

Section 50.7 also issued under Pub. L.95-601, sec. 10, 92 Stat. 2951 (42 U.S.C. 5851). Section SO.IO also issued under secs. 101, 185, 68 Stat.

955 as uiended (42 U.S.C. 2131, 2235), sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, S0.54(dd) and 50.103 also issued under sec.

108, 68 Stat. 939, as uiended (42 U.S.C. 2138). Sections 50.23, 50.35, 50.55, and 50.56 also issued under sec. 185, 68 Stat. 955 (42 U.S.C. 2235). Sections 50.33a, 50.55a and Appendix Q also issued under sec. 102, Pub. L.91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.34 and 50.54 also issued under sec.

204, 88 Stat. 1245 (42 U.S.C. 5844). Sections 50.58, 50.91 and 50.92 also issued under Pub. L.91-415, 96 Stat. 2073 (42 U.S.C. 2239). Section 50.78 also issued under sec. 122, 68 Stat. 939 (42 U.S.C. 2152). Sections 50.80 -

SO.Bl also issued under sec. 184, 68 Stat. 954, as amended (42 U.S.C. 2234).

Appendix Falso issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).

65

4. Section 50.2 is revised by adding in alphabetical order the definitions for Co111nitted dose eauivalent, Connitted effective dose eauiyaJent, Deep-dose eauivaJent, Exclusion area. Low population zone, Safetv-reJated structures. systems, and co11Qonents and Total effective dose eau1yalent, and revising the definition for Basic cOfflDonent (1)(111) to read as follows:

s 50.2 Definitions.

(1)

Basic component.

(111) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to those referred to in s50.34(a)(l) or sl00.11 of this chapter, as applicable.

CQ111111tted dose egu1valent means the dose equivalent to organs or tissues of reference that will be received from an intake of radioactive \

material by an individual during the 50-year period following the intake.

Comitted effective dose egu1valent is the sum of the products of the weighting factors applicable to each of the body organs or tissues that are irradiated and the c011111itted dose equivalent to these organs or tissues.

Deep-dose equivalent, which applies to external whole-body exposure, is the dose equivalent at a tissue depth of 1 cm (lOOOmg/car).

66

Exclusjon area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or removal of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with normal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area shall nornally be prohibited. In any event, residents shall be subject to ready removal in case of necessity. Activities unrelated to operation of the reactor may be permitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.

Low population zone means the area i11111ediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident.

These guides do not specify a permissible population density or total population within this zone because the situation may vary from case to case.

Whether a specific number of people can, for example, be evacuated from a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area.

67

Safety-related structures systems and components means those structures, systems, and components that are relied on to remain functional during and following design basis (postulated) events to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; and (3) The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth ins 50.34(a)(l) ors 100.11 of this chapter, as applicable.

• Total effective dose egyiyalent {TEDE) means the sum of the deep-dose equivalent (for external exposures) and the comitted effective dose equivalent (for internal exposures).

5. In sS0.8, paragraph {b) is revised to read as follows:

1 50.8 Information collection requirements: OIIB approval.

(b} The approved information collection requirements contained in this part appear in ssS0.30, 50.33, 50.33a, 50.34, 50.34a, 50.35, 50.36, 50.36a,

50.36b, 50.44, 50.46, 50.47, 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 50.62, 50.63, 50.64, 50.65, 50.66, 50.71, 50.72, 50.74, 50.75, 50.80, 50.82, 50.90, 50.91, 50.120, and Appendices A, B, E, G, H, I, J, K, M, N, 0, Q, R, and S to this part.

6. In s50.34, footnotes 6, 7, and 8 are redesignated as footnotes 8, 9 and 10 and paragraph (a)(l) 1s revised and paragraphs (a)(12),

• (b)(l0), and (b)(ll) are added to read as follows:

s 50.34 Contents of applications; technical 1nfomat1on.

(a) * * *

(1) Stationary power reactor applicants for a construction permit pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter who apply on or after [INSERT EFFECTIVE DATE OF THE FINAL RULE], shall comply with paragraph (a)(l)(ii) of this section. All other applicants for a construction permit pursuant to this part or a design certification or combined license pursuant to Part 52 of this chapter, shall COIIJ)ly with paragraph (a)(l)(i) of this section.

(i) A description and safety assessment of the site on which the facility is to be located, with appropriate attention to features affecting facility design. Special attention should be directed to the site evaluation factors identified in Part 100 of this chapter. The assessment must contain an 69

analysis and evaluation of the major structures, systems and components of the facility which bear significantly on the acceptability of the site under the site evaluation factors identified in Part 100 of this chapter, assuming that the facility will be operated at the ultimate power level which is contell'll)lated by the applicant. With respect to operation at the projected initial power level, the applicant is required to submit information prescribed in paragraphs (a)(2) through (a)(8) of this section, as well as the information required by this paragraph, in support of the application for a construction permit, or a design approval.

(ii) A description and safety assessment of the site and a safety assessment of the facility. It is expected that reactors will reflect through their design, construction and operation an extremely low probability for accidents that could result in the release of significant quantities of radioactive fission products. The following power reactor design characteristics and proposed operation will be taken into consideration by the C011111ission:

(A) Intended use of the reactor including the proposed maximum power level and the nature and inventory of contained radioactive materials; (B) The extent to which generally accepted engineering standards are applied to the design of the reactor; (C) The extent to which the reactor incorporates unique, unusual or enhanced safety features having a significant bearing on the probability or consequences of accidental release of radioactive materials; (D) The safety features that are to be engineered into the facility and those barriers that must be breached as a result of an accident before a 70

release of radioactive aaterial to the environment can occur. Special attention aust be directed to plant design features intended to mitigate the radiological consequences of accidents. In perfonning this assessment, an applicant shall assume a fission product release' from the core into the contain11ent assU11ing that the facility 1s operated at the ultimate power level contel'lplated. The applicant shall perfor11 an evaluation and analysis of the postulated fission product release, using the expected de110nstrable containaent leak rate and any fission product cleanup systems intended to mitigate the consequences of the accidents, together with applicable site characteristics, including site aeteorology, to evaluate the offsite radiological consequences. Site characteristics must comply with Part 100 of this chapter. The evaluation must determine that:

(l) An individual located at any point on the boundary of the exclusion area for any 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 25 re1111 total effective dose equivalent (TEDE).

(?) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its

• The fi111on product release aa111118d for thh evaluation should be hued upon a *Jor accident, hypothesized for purposn of site analysis or postulated frc. considerations of possible accidental events.

Such accidents have generally been asslllll!CI to result in substantial aeltdown of the core with subsequent release into the contaiment of appMtC1able quant1t1H of fhsion products.

' A whole body dote of 25 ra has been 11tated tn correspond n1.1111erically to the once in a 11fet1111e accidental or 11119rgency dose for radiation workers which. according to NCRP recc.aendations at the t111e could be disregarded 1n the detersinatlon 01* their radiation exposure status (see NBS Handbook. 69 dated June 5, 1959). However, its use 11 not intended to illl)ly that this nUllber constitutes an acceptable limit for an emrgency dose to the public under accident conditions. Rather, this dose value has been set forth in this section as a reference value, which can be used In the evaluation of plant design futures with respect to postulated reactor accidents, 1n order to assure that such designs provide ass11rance of low risk of public exposure to radiation, in the event of such accidents.

71

passage) would not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE).

(E) With respect to operation at the projected initial power level, the applicant is required to subllit infonntion prescribed in paragraphs (a)(2) through (a)(B) of this section, as well as the infomation required by this paragraph, in support of the application for a construction permit, or a design approval.

(12) On or after [INSERT EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for a construction permit pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter, as partial conformance to General Design Criterion 2 of Appendix A to this part, shall c011ply with the earthquake engineering criteria in Appendix S to this part.

(b) * * *

(10) On or after [INSERT EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for an operating license pursuant to this part, or a design certification or combined license pursuant to Part 52 of this chapter, as partial conformance to General Design Criterion 2 of Appendix A to this part, shall c0111ply with the earthquake engineering criteria of Appendix S to this part. However, for those operating license applicants and holders whose construction permit was issued prior to [INSERT EFFECTIVE DATE 72

OF THE FINAL RULE], the e:rthquake engineering criteria in Section VI of Appendix A to Part 100 of this chapter continues to apply.

(11) On or after [INSERT EFFECTIVE DATE OF THE FINAL RULE], stationary power reactor applicants who apply for an operating license pursuant to this Part, or a COlllbined license pursuant to Part 52 of this chapter, shall provide a description a~d safety assessment of the site and of the facility as 1n sS0.34{a){l}(ii) of this part. However, for either an operating license

• applicant or holder whose construction permit was issued prior to [INSERT EFFECTIVE DATE OF THE FINAL RULE], the reactor site criteria in Part 100 of this chapter and the seismic and geologic siting criteria in Appendix A to Part 100 of this chapter continues to apply.

7 In sS0.49, paragraph (b)(l) is revised to read as follows:

s 50.49 Envirol"IIAlntal qua11f1cat1on of electric equipment important to safety for nuclear power plants.

(b) * *

• 73

(1) Safety-related electric equipinent. 3 (1) This equipaent is that relied upon to remain functional during and following design basis events to ensure --

(A) The integrity of the reactor coolant pressure boundary; (8) The capability to shut down the reactor and maintain it in a safe shutdown condition; and (C) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guidelines in sS0.34(a)(l) or sl00.11 of this chapter, as applicable.

(11) Design basis events are defined as conditions of normal operation, including anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant must be designed to ensure functions (b)(l)(i)(A) through (C) of this section.

8. In sS0.54, paragraph (ff) is added to read as follows:

150.54 Conditions of licenses.

(ff) For licensees of nuclear rower plants that have implemented the a

Safety-related electric equipment 1s referred t0 aa "Claaa lE" equiplllBl'lt in IEEE 323-1974. Copies of this standard aay be obtained fn::a the Institute of Electrical and Electronics Engineers, Inc., 3-iS East 47th Street, New York, NY 10017.

74

earthquake engineering criteria in Appendix S to this part, plant shutdown is required as provided in Paragraph IV(a)(3) of Appendix S. Prior to resuming operations, the licensee shall demonstrate to the Conrnission that no functional da~age has occurred to those features necessary for continued operation without undue risk to the health and safety of the public and the licensing basis is maintained.

9. In s50.65, paragraph (b)(l) is revised to read as follows:

s 50.65 Requirements for monitoring the effectiveness of maintenance at nuclear power plants (b) * * *

(1) Safety related structures, systems, or components that are relied upon to remain functional during and following design basis events to ensure the integrity of the reactor coolant pressure boundary, the capability to shut down the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in s50.34(a)(l) or sl00.11 of this chapter, as applicable.

75

10. Appendix S to Part 50 is added to read as follows:

APPENDIX S TO PART 50 - EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS General Information This appendix applies to applicants for a design certification or combined license pursuant to Part 52 of this chapter or a construction permit or operating license pursuant to Part 50 of this chapter on or after [INSERT EFFECTIVE DATE OF THE FINAL RULE]. However, for either an operating license applicant or holder whose construction permit was issued prior to [INSERT EFFECTIVE DATE OF THE FINAL RULE], the earthquake engineering criteria in Section VI of Appendix A to 10 CFR Part 100 continues to apply.

I. Introduction Each applicant for a construction permit, operating license, design certification, or cOllbined license is required by s50.34(a)(l2), (b)(lO), and General Design Criterion 2 of Appendix A to this Part to design nuclear power plant structures, systems, and components important to safety to withstand the effects of natural phenomena, such as earthquakes, without loss of capability to perform their safety functions. Also, as specified ins S0.54(ff), nuclear power plants that have implemented the earthquake engineering criteria 76

described herein must shut down if the criteria in Paragraph IV(a)(3) of this appendix are exceeded.

These criteria impleaent General Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes.

II. Scope The evaluations described in this Qppendix are within the scope of investigations permitted by sSO.IO(c)(l).

Ill. Definitions As used in these criteria:

COIRbjned license means a combined construction permit and operating license with conditions for a nuclear power facility issued pursuant to Subpart C of Part 52 of this chapter.

Design Certificatjon means a Co11111ission approval, issued pursuant to Subpart B of Part 52 of this chapter, of a standard design for a nuclear power facility. A design so approved may be referred to as a *certified standard design.*

77

The Operating Basis Earthquake Ground Motion COBE} is the vibratory ground motion for which those features of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public will remain functional. The Operating Basis Earthquake Ground Motion is only associated with plant shutdown and inspection unless specifically selected by the applicant as a design input.

A response spectrum is a plot of the maximum responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports.

The Safe Shutdown Earthquake Ground Motion (SSE) is the vibratory ground 1n0tion for which certain structures, systems, and components must be designed to remain functional.

The structures, systems, and components required to withstand the effects of the Safe Shutdown Earthquake Ground Motion or surface deformation are those necessary to assure:

(1) The integrity of the reactor coolant pressure boundary; (2) The capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of sS0.34(a)(l).

78

Surface deformation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface deformation is associated with earthquake processes.

IV. Application To Engineering Design The following are pursuant to the seismic and geologic desig~ basis requireaents of sl00.23 of this chapter:

(a) Vibratory Ground Notion.

(1) Safe Shutdown Earthquake &round Notion. The Safe Shutdown Earthquake Ground Motion must be characterized by free-field ground motion response spectra at the free ground surface. In view of the limited data available on vibratory ground 110tions of strong earthquakes, it usually will be appropriate that the design response spectra be smoothed spectra. The horizontal component of the Safe Shutdown Earthquake Ground Motion in the free-field at the foundation 1evel of the structures must be an appropriate response spectrum with a peak ground acceleration of at least O.lg.

The nuclear power plant must be designed so that, if the Safe Shutdown Earthquake Ground Motion occurs, certain structures, systems, and components will reaain functional and within applicable stress, strain, and defonnation limits. In addition to seismic loads, applicable concurrent normal operating, functional, and accident-induced loads must be taken into account in the design of these safety-related structures, systems, and components. The design of the nuclear power plant IRllst also take into account the possible effects of 79

the Safe Shutdown Earthquake Ground Motior, on the facility foundations by ground disruption, such as fissuring, lateral spreads, differential settleaent, liquefaction, and landsliding, as required in sl00.23 of this chapter.

The required safety functions of structures, systems, and components must be assured during and after the vibratory ground motion associated with the Safe Shutdown Earthquake Ground Motion through design, testing, or qualification methods.

The evaluation 11Ust take into account soil-structure interaction effects and the expected duration of vibratory motion. It is permissible to design for strain limits in excess of yield strain in some of these safety-related structures, systems, anq components during the Safe Shutdown Earthquake Ground Motion and under the postulated concurrent loads, provided the necessary safety functions are maintained.

(2) Operating Basis Earthquake &round Notion.

(i) The Operating Basis Earthquake Ground Motion must be characterized by response spectra. The value of the Operating Basis Earthquake Ground Motion must be set to one of the following choices:

(A) One-third or less of the Safe Shutdown Earthquake Ground Motion design response spectra. The requireaents associated with this Operating Basis Earthquake Ground Motion in Paragraph (a)(2)(i)(B)(I) can be satisfied without the applicant performing explicit response or design analyses, or (B) A value greater than one-third of the Safe Shutdown Earthquake Ground Motion design response spectra. Analysis and design must be performed to demonstrate that the requirements associated with this Operating Basis 80

Earthquake Ground Motion ~n Paragraph (a)(2)(i)(B)(I) are satisfied. The design 11Ust take into account soil-structure interaction effects and the duration of vibratory ground motion.

(I) When subjected to the effects of the Operating Basis Earthquake Ground Motion in combination with normal operating loads, all structures, systems, and c01Dponents of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public 111.1st remain functional and within applicable stress, strain, and deformation limits.

(3) Required Plant Shutdown. If vibratory ground motion exceeding that of the Operating Basis Earthquake Ground Motion or if significant plant damage occurs, the licensee must shut down the nuclear power plant. If systems, structures, or components necessary for the safe shutdown of the nuclear power plant are not available after the occurrence of the Operating Basis Earthquake Ground Motion, the licensee 11Ust consult with the Connission and must propose a plan for the tiaely, safe shutdown of the nuclear power plant. Prior to resU11ing operations, the licensee 111.1st de110nstrate to the Comission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public and the licensing basis is aaintained.

(4) Required Se1s1c Instrl1Nntat1on. Suitable instrumentation must be provided so that the seismic response of nuclear power plant features important to safety can be evaluated promptly after an earthquake.

(b) surface Defonntion. The potential for surface deformation must be taken into account in the design of the nuclear power plant by providing reasonable assurance that in the event of deformation, certain structures, 81

systems, and COflf)Onents will remain functional. In addition to surface deformation induced loads, the design of safety features must take into account seismic loads and applicable concurrent functional and accident-induced loads. The design provisions for surface deformation must be based on its postulated occurrence 1n any direction and azimuth and under any part of the nuclear power plant, unless evidence indicates this assumption is not appropriate, and aust take into account the estimated rate at which the surface deformation ay occur.

(c) Se1Slt1ca11y Induced Floods and Water Waves and Other Design Conditions. Seis ically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to sl00.23 of this chapter must be taken into account in the design of the nuclear power plant so as to prevent undue risk to the health and safety of the public.

PART 52- EARLY SITE PERMITS; STANDARD DESIGN CERTIFICATIONS; AtlJ COMBINED LICENSES FOR NUCLEAR POWER PLANTS

11. The authority citation for Part 52 continues to read as follows:

AUTHORITY: Secs. 103, 104, 161, 182, 183, 186, 189, 68 Stat. 936, 948, 953, 954, 955, 956, as aaended, sec. 234, 83 Stat. 1244, as amended (42 U.S.C.

2133, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, 1246, as amended (42 U.S.C. 5841, 5842, 5846).

82

12. In s52.17, the introductory text of paragraph (a)(l) and paragraph (a)(l)(vi) are revised to read as follows:

162.17 Contents of applications.

(a)(l) The application must contain the information required bys 50.33(a)-(d), the information required bys 50.34 (a)(12) and (b)(l0), and to the extent approval of eaergency plans is sought under paragraph (b)(2)(ii) of this section, the infonnation required bys 50.33 (g) and (j), ands 50.34 (b)(6)(v). The application 11Ust also contain a description and safety assess11ent of the site on which the facility is to be located. The assessment must contain an analysis and evaluation of the ajor structures, systems~ and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified ins 50.34(a)(l) of this chapter. Site characteristics must comply with Part 100 of this chapter. In addition, the application should describe the following:

(vi) The seismic, aeteorological, hydrologic, and geologic characteristics of the proposed site; PART 54 - REQUIREMENTS FOR RENEWAL OF OPERATING LICENSES FOR NUCLEAR POWER PLANTS 13 The authority citation for Part 54 continues to read as follows:

83

AUTHORITY: Secs. 102, 103, 104, 161, 181, 182, 183, 186, 189, 68 Stat.

936, 937, 938, 948, 953, 954, 955, as amended, sec. 234, 83 Stat. 1244, as 1111ended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); secs. 201, 202, 206, 88 Stat. 1242, 1244, as amended (42 U.S.C. 5841, 5842).

14 In s54.4, paragraph (a)(l)(111) 1s revised to read as follows:

154.4 Scope.

(a) * * *

(1) * * *

(111) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposure comparable to the guidelines in s50.34(a)(l) or sl00.11 of this chapter, as applicable.

PM.T 100 - REACTOR SITE CRITERIA

15. The authority citation for Part 100 continues to read as follows:

AUTHORITY: Secs. 103, 104, 161, 182, 68 Stat. 936, 937, 948, 953, as 84

aaended (42 U.S.C. 2133, 2134, 2201, 2232); sec. 201, as amended, 202, 88 Stat. 1242, as amended, 1244 (42 U.S.C. 5841, 5842).

16. The table of contents for Part 100 is revised to read as follows:

PART 100 - REACTOR SITE CRITERIA Sec.

100.1 Purpose.

100.2 Scope.

100.3 Definitions.

100.4 Comunications.

100.8 Information collection requirements: 0MB approval.

Subpart A - Evaluation Factors for Stationary Power Reactor Site Applications Before [EFFECTIVE DATE OF THE FINAL RULE] and for Testing Reactors.

100.10 Factors to be considered when evaluating sites.

100.11 Determination of exclusion area, low population zone, and population center distance.

subpart B - Evaluation Factors for Stationary Power Reactor Site Applications on or After [EFFECTIVE DATE OF THE FINAL RULE].

100.20 Factors to be considered when evaluating sites.

85

100.21 Non-seismic site criteria.

100.23 Geologic and seismic siting criteria.

APPENDIX A to Part 100 - Seismic and Geologic Siting Criteria for Nuclear Power Plants.

17. Section 100.1 is revised to read as follows:

s 100.1 Purpose.

(a) The purpose of this part is to establish approval requirements for proposed sites for stationary power and testing reactors subject to Part 50 or Part 52 of this chapter.

(b) There exists a substantial base of knowledge regarding power reactor siting, design, construction and operation. This base reflects that the primary factors that determine public health and safety are the reactor design, construction and operation.

(c) Siting factors and criteria are illf)ortant in assuring that radiological doses from nonnal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, that site characteristics are such that adequate security measures to protect the plant can be developed, and that physical characteristics unique to the proposed site that could pose a significant impediment to the development of emergency plans are identified.

86

(d) This approach incorporates the appropriate standards and criteria for approval of stationary power and testing reactor sites. The Con111ission intends to carry out a traditional defense-in-depth approach with regard to reactor siting to ensure public safety. Siting away from densely populated centers has been and will continue to be an illll)ortant factor in evaluating applications for site approval.

18. Section 100.2 is revised to read as follows:

I 100.2 Scopa.

The siting requirements contained in this part apply to applications for site approval for the purpose of constructing and operating stationary power and testing reactors pursuant to the provisions of Parts 50 or 52 of this chapter.

19. Section 100.3 is revised to read as follows:

s 100.3 Definitions.

As used in this part:

COllbined license aeans a combined construction permit and operating license with conditions for a nt*clear power facility issued pursuant to Subpart C of Part 52 of this chapter.

87

Early Site Perajt means a Comatssion approval, issued pursuant to subpart A of Part 52 of this chapter, for a site or sites for one or more nuclear power facilities.

Exclusion area means that area surrounding the reactor, in which the reactor licensee has the authority to determine all activities including exclusion or re110val of personnel and property from the area. This area may be traversed by a highway, railroad, or waterway, provided these are not so close to the facility as to interfere with nomal operations of the facility and provided appropriate and effective arrangements are made to control traffic on the highway, railroad, or waterway, in case of emergency, to protect the public health and safety. Residence within the exclusion area shall normally be prohibited. In any event, r~sidents shall be subject to ready r&1110val in case of necessity. Activities unrelated to operation of the reactor may be peraitted in an exclusion area under appropriate limitations, provided that no significant hazards to the public health and safety will result.

Low population zone means the area t11111ediately surrounding the exclusion area which contains residents, the total number and density of which are such that there is a reasonable probability that appropriate protective measures could be taken in their behalf in the event of a serious accident. These guides do not specify a pennissible population density or total population w;thin this zone because the situation may vary from case to case. Whether a specific number of people can, for example, be evacuated fr011 a specific area, or instructed to take shelter, on a timely basis will depend on many factors such as location, number and size of highways, scope and extent of advance planning, and actual distribution of residents within the area.

Population center di~tance 111eans the distance from the reactor to the nearest boundary of a densely populated center containing more than about 25,000 residents.

Power reactor means a nuclear reactor of a type described in ssS0.2l(b) or 50.22 of this chapter designed to produce electrical or heat energy.

A Response spectrum is a plot of the maxiawa responses (acceleration, velocity, or displacement) of idealized single-degree-of-freedom oscillators as a function of the natural frequencies of the oscillators for a given

- daJRp1ng value. The response spectrum is calculated for a specified vibratory 110tion input at the oscillators' supports.

The Safe Shutdown Earthquake Ground Motion is the vibratory ground motion for which certain structures, systems, and components must be designed pursuant to Appendix S to Part 50 of this chapter to remain functional.

Surface defonnation is distortion of geologic strata at or near the ground surface by the processes of folding or faulting as a result of various earth forces. Tectonic surface defonaation is associated with earthquake processes.

Testing reactor means a testing facility as defined in sS0.2 of this chapter.

20. Section 100.4 is added to read as follows:

sl00.4 Coaiunications.

Except where otherwise specified in this part, all correspondence, reports, applications, and other written cOR111unications submitted pursuant to 89

10 CFR Part 100 should be addressed to the U.S. Nuclear Regulatory Connission, ATTN: DocU11ent Control Desk, Washington, DC 20555-0001, and copies sent to the appropriate Regional Office and Resident Inspector. Comunications and reports may be delivered in person at the Con.ission's offices at 2120 L Street, NW., Washington, DC, or at 11555 Rockville Pike, Rockville, Maryland.

21. Section 100.8 is revised to read as follows:

s 100.8 Inforut1on collection requirements: ONB approval.

(a) The Nuclear Regulatory Coranission has submitted the information collection requirements contained in this part to the Office of Management and Budget (0MB) for approval as required by the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.). 0MB has approved the information collection requirements contained in this part under control number 3150--0093.

(b) The approved information collection requirements contained in this part appear in sl00.23 and Appendix A.

22. A heading for Subpart A is added directly before sl00.10 to read as follows:

Subpart A- Evaluation Factors for Stationary Power Reactor Site Applications before [EFFECTIVE DATE OF THIS REGULATION] and for Testing Reactors.

90

23. Sections 100.20, 100.21 and 100.23 are added to Subpart B to read as follows:

Subpart B - Evaluation Factors for Stationary Power Reactor Site Applications on or After [EFFECTIVE DATE OF THE FINAL RULE].

1100.20 Factors to be considered when evaluating sites.

The C011111ission will take the following factors into consideration in determining the acceptability of a site for a stationary power reactor:

(a) Population density and use characteristics of the site environs, including the exclusion area, the population distribution, and site-related characteristics IDUSt be evaluated to deter11ine whether individual as well as societal risk of potential plant accidents is low, and that physi_cal characteristics unique to the proposed site that could pose a significant i111pediment to the developaent of emergency plans are identified.

(b) The nature and proximity of an-related hazards (e.g., airports, dams, transportation routes, military and chemical facilities) 111.1st be evaluated to establish site paraaeters for use in determining whether a plant design can acconnodate co111110nly occurring hazards, and whether the risk of other hazards is very low.

(c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology.

91

(1) Section 100.23, *Geologic and seismic siting factors,* describes the criteria and nature of investigations required to obtain the geologic and seismic data necessary to determine the suitability of the proposed site and the plant design bases.

(2) Meteorological characteristics of the site that are necessary for safety analysis or that may have an impact upon plant design (such as maximUIII probable wind speed and precipitation) aust be identified and characterized.

(3) Factors important to hydrological radionuclide transport (such

• as soil, sediaent, and rock characteristics, adsorption and retention coefficients, ground water velocity, and distances to the nearest surface body of water) ust be obtained from on-site measurements. The maximum probable flood along with the potential for seismically induced floods discussed in sl00.23 (d)(3) of this part must be estimated using historical data.

s 100.21 Non-seismic siting criteria.

Applications for site approval for connercial power reactors shall demonstrate that the proposed site meets the following criteria:

(a) Every site must have an exclusion area and a low population zone, as defined in sl00.3; (b) The population center distance, as defined in sl00.3, must be at least one and one-third times the distance from the reactor to the outer boundary of the low population zone. In applying this guide, the boundary of 92

the population center shall be determined upon consideration of population distribution. Political boundaries are not controlling in the application of this guide; (c) Site at110spheric dispersion characteristics must be evaluated and dispersion parameters established such that:

(1) Radiological effluent release li its associated with normal operation from the type of facility proposed to be located at the site can be met for any individual located offsite; and (2) Radiological dose consequences of postulated accidents shall meet the criteria set forth.in s50.34(a)(l) of this chapter for the type of facility proposed to be located at the site; (d) The physical characteristics of the site, including meteorology, geology, seismology, and hydrology must be evaluated and site parameters established such that potential threats frORI such physical characteristics will pose no undue risk to the type of facility proposed to be located at the site; (e) Potential hazards associated with nearby transportation routes, industrial and military facilities must be evaluated and site parameters established such that potential hazards from such routes and facilities will pose no undue risk to the type of facility proposed to be located at the site; 93

(f) Site characteristics must be such that adequate security plans and measures can be developed; (g) Physical characteristics unique to the proposed site that could pose a significant inipedi11ent to the development of emergency plans must be identified; (h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in deteraining the acceptability of a particular site located away from a very densely populated center but not in an area of low density, consideration will be given to safety, environmental, econ011ic, or other factors, which may result in the site being found acceptable3

• s 100.23 Geologic and seismic siting factors.

This section sets forth the principal geologic and seis ic considerations that guide the C0111Rission in its evaluation of the suitability of a proposed site and adequacy of the design bases established in consideration of the geologic and seis~ic characteristics of the proposed site, such that, there is a reasonable assurance that a nuclear power plant can be constructed and operated at the proposed site without undue risk to the

>> Exanplea of these factors include, but are not li*ited to, such factors as the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better ratl and highway access, 1horter transa1 s1on line requirements, or leH environmental 1111p4ct on undeveloped areas, wetlands or endangered species, etc. Some of these factors are included in, or i11'4)11ct, the other criteria included 1n this section.

94

health and safety of the public. Applications to engineering design are contained in Appendix S to Part SO of this chapter.

(a) App11cab11tty. The requirements in paragraphs (c) and (d) of this section apply to applicants for an early site permit or combined license pursuant to Part 52 of this chapter, or a construction permit or operating license for a nuclear power plant pursuant to Part 50 of this chapter on or after [INSERT EFFECTIVE DATE OF THE FINAL RULE]. However, for either an operating license applicant or holder whose construction permit was issued prior to [INSERT EFFECTIVE DATE OF THE FINAL RULE], the seismic and geologic siting criteria in Appendix A to Part 100 of this chapter continues to apply.

(b) Comencement of construction. The investigations required in paragraph (c) of this section are within the scope of investigations permitted bys 50.lO(c)(l) of this chapter.

(c) 6eolog1ca1, se1s1101og1ca1, and engineering character1st1cs. The geological, seismological, and engineering characteristics of a site and its environs must be investigated in sufficient scope and detail to permit an adequate evaluation of the proposed site, to provide sufficient information to support evaluations perfonned to arrive at estimates of the Safe Shutdown Earthquake Ground Motion, and to permit adequate engineering solutions to actual or potential geologic and seismic effects at the proposed site. The size of the region to be investigated and the type of data pertinent to the investigations 11t.1st be determined based on the nature of the region surrounding the proposed site. Data on the vibratory ground motion, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault geometry and slip rates, site foundation material, and seismically induced floods and water waves IIUSt be obtained by reviewing pertinent 95

literature and carrying out field investigations. However, each applicant shall investigate all geologic and seismic factors (for example, volcanic activity) that uy affect the design and operation of the proposed nuclear power plant irrespective of whether such factors are explicitly included in this section.

(d) Geologic and se1sa1c siting factors. The geologic and seismic siting factors considered for design raust include a determination of the Safe I

Shutdown Earthquake Ground Motion for the site, the potential for surface tectonic and nontectonic defonwations, the design bases for seismically induced floods and water waves, and other design conditions as stated in paragraph (d)(4) of this section.

(1) Detemination of the Safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion for the site is characterized by both horizontal and vertical free-field ground motion response spectra at the free ground surface. The Safe Shutdown Earthquake Ground Motion for the site is detenained considering the results of the investigations required by paragraph (c) of th~s section. Uncertainties are inherent in such estimates. These uncertainties lllUSt be addressed through an appropriate analysis, such as a probabilistic seisaic hazard analysis or suitable sensitivity analyses.

Paragraph IV(a)(l) of Appendix S to Part 50 of this chapter defines the Mini1111m Safe Shutdown Earthquake Ground Motion for design.

(2) Determination of the potential for surface tectonic and nontectonic defonnations. Sufficient geological, seismological, and geophysical data must be provided to clearly establish whether there is a potential for surface deformation.

96

(3) Detennination of design bases for seismically induced floods and water waves. The size of seismically induced floods and water waves that could affect a site from either locally or distantly generated seismic activity must be determined.

{4} Detenninat1on of siting factors for other design conditions. Siting factors fo.r other design conditions that must be evaluated include soil and rock stability, liquefaction potential, natural and artificial slope stability, cooling water supply, and remote safety-related structure siting.

Each applicant shall evaluate all siting factors and potential causes of failure, such as, the physical properties of the materials underlying the site, ground disruption, and the effects of vibratory gro~nd motion that may affect the design and operation of the proposed nuclear power plant.

'} 11.I Dated at Rockville, Maryland, this _?<_-day of December, 1996.

e For the Nuclear Regulatory Commission.

Cor1111ission.

97

NEWMAN & HOLTZINGER, P. C.

ATTORNEYS AT LAW 1615 L STREET, N.W.

• 93 OCT -4 P3 :44 WASHINGTON, D . C . 20036*5610 TELEPHONE: (202) 955-6600 FAX: (202) 872-0581 October 1, 1993 Mr. Samuel J. Chilk Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Docketing and Service Branch Re: Proposed Rule on Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition for Rulemaking From Free Environment, Inc. et al. (57 Fed.

Reg. 47,802 (October 20, 1992))

Dear Mr. Chilk:

On June 1, 1993, the law firm of Newman & Holtzinger, P.C., on behalf of its clients in its International Siting Group

• (ISG), submitted comments on the Nuclear Regulatory Commission's proposed ru le , "Proposed Rule on Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition f or Rulemaking From Free Environment, Inc. et al." (57 Fed. Reg. 47,802 (October 20, 1992)). On August 3, 1993, the Commission met with members of the NRC Staff to discuss comments received on the proposed rule, in particular, comments from the international community, and alternatives to the proposed rule, especially in order to address such comments. On August 12, 1993, the Secretary to the Commission issued a Staff Requirements Memorandum (SRM),

formalizing t he guidance which the Commission provided to the Staff during the August 3rd meeting concerning consideration of alternatives.

In light of these developments, we are pleased to supplement the earlier comments of the I SG on the proposed rule.

Specifically, we address each of the issues identified in the SRM, giving the position of ISG Members on the issue. Based on that position, we discuss the acceptabi l ity or unacceptability of the various alternatives.

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NEWMAN & HOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 2 At the outset, it is important to emphasize that the ISG's position, as stated in our original submission, has not changed. ISG Members remain of the view that the Commission should withdraw the proposed revisions to the siting criteria and terminate the rulemaking proceeding for the reasons expressed in the original submission. However, based on discussion during the August 3rd Commission meeting and in the August 12th SRM concerning consideration of alternatives, the value of providing supplemental views on alternative courses of action is.

recognized. For the reasons discussed below, the following alternatives appear acceptable for further consideration, should the Commission decide not to terminate the present rulemaking proc~eding:

(1) Leave Part 100 and Reg. Guide 4.7 as they are, except replace the reference to the TIO 14844 source terms with reference to more realistic source terms, which reflect technological advances and improved understandings of accident progression and risk.

(2) State general objectives which conform to stated risk objectives, such as the Safety Goals. This alternative would. be potentially acceptable provided that a relationship between the general objectives and risk objectives could be established. In that case, general objectives would be set in terms of their risk reduction potential.

(3) Defer consideration of any changes to the siting regulation until after the source term update and severe accident rulemakings are completed. Specifically, complete the source term update program, followed by establishment of severe accident requirements through rulemaking, followed by consideration of revisions to the siting regulations.

The following alternatives are unacceptable:

(1) The proposed rule and any alternatives which would eliminate dose calculations from the site evaluation process, including an alternative which would retain the form of the proposed rule, but put the numbers in a regulatory guide or an alternative which screens out or favors sites on the basis of demographic considerations alone.

(2) Alternatives which establish a hierarchy among the 14 siting parameters or otherwise diminish the flexibility necessary to select sites which are among the best reasonably to

NEWMAN & HoLTZINoER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 3 be found with respect to ensuring adequate protection of the public health and safety and the environment.

DISCUSSION Issue No. 1: The extent to which source terms can be decoupled from the siting criteria in view of technological advancements.

We understand this issue to mean that dose calculations would no longer be used in judging the suitability of a partioular site to serve as the location for a nuclear power plant of a particular design. The NRC Staff's concern is that for modern nuclear power plant designs, the source terms are such that use of dose calculations could compromise achievement of the goal of siting nuclear power plants remotely to the extent practicable. The concern was addressed in the proposed rule by eliminating dose calculations as a basis for making decisions about particular nuclear plant sites and plant-site combinations.

Dose calculations should continue to be used in determining the suitability of a particular site and plant-site combination. Nuclear power plant siting involves consideration of 14 different siting parameters and a weighing of the significance of these factors for a particular site. Dose calculations permit an overall figure of merit to be developed for each site and plant-site combination taking into account these factors. This figure of merit is important in integrating these factors, judging suitability and in choosing a preferred site. Indeed, even 10 C.F.R. Part 52, which provides for' the separation of siting decisions from design decisions, requires the specification of plant parameters in conjunction with early site approval pursuant to Subpart A of Part 52 and specification of site parameters in conjunction with design approval pursuant to Subpart B. Performance of dose calculations is an important component in the review process for each kind of approval.

As discussed in our original submission, elimination of dose calculations is not necessary to achieve remote siting. As analysis of U.S. nuclear power plant siting data clearly demonstrates, present Part 100 and implementing practice have achieved remote siting. Elimination of dose calculations would not make this more achievable and would eliminate a useful and accepted tool for judging, overall suitability of the site and the plant-site combination.

NEWMAN & liOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 4 Also, el.:µnination of dose calculations from consideration on a site-specific basis would not necessarily de-couple source terms from siting decisions. Dose calculations would likely be used in selecting the siting criteria to be made part of the regulations since they would provide part of the basis for judging how application of the criteria could be expected to reduce the radiological risk. Moreover, Part 52 already substantially decouples siting from design by emphasizing selection of robust designs and robust sites.

For these reasons, ISG Members find unacceptable both the proposed rule and such alternatives as would eliminate dose calculations from the site evaluation process.

Issue No. 2z Technical and safety-related basis for siting criteria as opposed to what the U.S. can accommodate.

The Commission's reactor siting regulations in Part 100 were intended to ensure adequate protection of the public health and safety in the siting of nuclear power plants. The Commission's regulations have achieved their intended purpose as analysis of nuclear power plant siting decisions clearly demonstrates. Therefore, change in the regulations is not necessary to achieve an adequate level of safety.

When safety improvements in nuclear power plants are taken into account, it is difficult to find a technical and safety basis for making the Commission's regulations more stringent. The safety risk simply does not justify greater stringency.

Severe accident risk and sabotage risk have been raised as potential justifications for more rigorous prescription of remote siting requirements in the Commission's regulations. With respect to severe accident risk, if the Commission wishes to justify more stringent siting requirements on the basis of severe accident risk, it seems preferable to change the order in which the source term is updated, severe accident requirements are developed and the siting regulations are revised. Presently, the order is: (1) revise the siting rule; (2) update the source terms; and (3) establish severe accident requirements. A more logical way to proceed would be toz (1) update the source terms; (2) establish severe accident requirements; and only then (3) consider revision of the siting requirements on the basis of severe accident risk.

NEWMAN & HOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 5 If the risk from sabotage is to serve as th~ basis for the revised rule, the basis should be made explicit and the relationship between the requirements and reduction in sabotage risk explained. Again, the consideration of more restrictive requirements should address the risk, how well the present requirements have contained that risk, and how the requirements under consideration would reduce that risk.

Taking the foregoing into account, an alternative which focused on update of the source terms in the nearer term, establishment of severe accident requirements in the middle term and consideration of revisions to the siting rule only after completion of the other two efforts would be acceptable.

Similarly, ISG Members would find it acceptable to evaluate whether significant reduction in the risk from sabotage can be achieved by imposing more stringent siting requirements than those already in place in Part 100. Only if significant reduction would result, should the Commission consider revising Part 100. '

Issue Ho. 3: Extent to which proposed reactor site criteria reflect concerns of potential users in other countries.

Remote siting of nuclear power plants to the extent practicable is the internationally accepted siting norm. ISG Member countries, as well as the International Atomic Energy Agency (IAEA) and other countries in the world, have incorporated this norm into their regulatory regimes and follow it in their siting practices.

Siting involves the balancing of 14 siting parameters.

International siting practices do not impose a hierarchy on these parameters, but embody flexibility to ensure appropriate weighing in a manner which provides for adequate protection of the public health and safety and the environment, including the capability to take appropriate emergency action in the event of a radiological emergency.

The international community is opposed t*o adoption of siting requirements which diminish the flexibility necessary to make reasoned siting decisions. U.S. safety standards cannot be and are not ignored elsewhere in the world. If the United States adopts more restrictive siting requirements on the basis that safety requires such stringency, then other countries must consider what the U.S. has done. If safety is implicated in the

NEWMAN & HOLTZINGER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 6 more stringent requirements and there is not a substantial increase in safety, countries elsewhere in the world must still consider what the U.S. has done and justify to their citizenry why they are not doing the same.

The proposed changes to the siting regulations take away flexibility without significant -- or even any -- increase in safety. Therefore, the alternative is unacceptable to ISG Members. Similarly, any alternative which diminishes flexibility in siting without an attendant and significant increase in safety is unacceptable. Among the alternatives described at the August 3rd meeting, unacceptable alternatives would include: (1) the proposed rule; (2) keeping the form of the proposed rule, but changing the numbers to reflect more realistic source terms and technological improvements in the design; and (3) putting numbers in the rule as limits such that if a site is within these limits, no further justification would be necessary.

Issue No. 41 Pros and cons of less prescriptive revisions to Part 100 than those issued for public comment.

During the August 3rd meeting, the NRC Staff identified several less prescriptive revisions to the proposed revisions to Part 100. These were: (1) Leave Part 100 and Reg. Guide 4.7 as they are, except replace the reference to the TID 14844 source terms with reference to more realistic source terms, which reflect technological advances and improved understandings of accident progression and risk; (2) State general objectives for an exclusion area and population density in a rule, but keep any numbers in Reg. Guide 4.7; and, potentially, (3) Defer further consideration of any changes to the siting criteria until after the source term update and severe accident rulemakings are completed. Specifically, complete source term update program, followed by establishment of severe accident requirements through rulemaking, followed by consideration of revisions to the siting regulations. A fourth less prescriptive alternative would be one based on conformance to stated risk objectives, such as the Safety Goal.

(1) Leave Part 100 and Reg. Guide 4.7 as they are, except replace the reference to the TID 14844 source terms with reference to more realistic source terms, which reflect technological_advances and improved understandings of accident progression and risk.

NEWMA.N & HoLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 7 This alternative would be acceptable because it would use realistic source terms which reflect the risk profiles of the new reactor designs. As such, it would provide a consistent regulatory regime for eva~uating new reactor designs, the sites on which they would be placed and the plant-site combination, as discussed above under Issue No. 1. Additionally, it would avoid establishing a hierarchy of siting parameters, with the attendant disadvantages discussed above under Issue No. 3. It would also retain remote siting as part of regulatory policy and practice.

Some might argue that use of realistic source terms would compromise remote siting. This will not happen. Remote siting +/-s universally accepted and embodied in U.S. practice.

Analysis of U.S. siting data supports this position. In instances where the Commission has believed a candidate site is too close to a population center, alternative sites have been found. There is no reason to expect a different outcome in the future.

Some also might argue that this alternative would continue past difficulties encountered in the licensing process concerning nuclear power plant siting. As long as the public has questions about the use of nuclear power and the licensing forum remains available to raise such questions, contentions will be raised concerning the location of nuclear power plants.

Inappropriately restrictive siting requirements without a safety basis, such as the proposed changes to Part 100, add to public concern by suggesting that such restrictions are necessary to ensure adequate protection of the public health and safety and the environment.

(2) & (4) State general objectives for an exclusion area and population density in a rule, but keep any numbers in Reg. Guide 4.7., basing such objectives on conformance to stated risk objectives, such as the Safety Goal.

A statement-of general objectives which conformed to stated risk objectives, such as the Safety Goals, would be potentially acceptable provided that a relationship between the general objectives and risk objectives could be established. In that case, general objectives would be set in terms of their risk reduction potential. The lack of a link between risk reduction and the proposed revisions to the siting rule was one of the major defects in the proposed revisions. In fact, review of the proposed revisions in light of the Commission's decision

NEWMAN & HOLTZINGER, P.C.

Mr. Samuel J. Chilk October 1, 1993 Page 8 framework established that there was no safety basis or risk reduction potential associated with the proposed revisions.

Stating general objectives as a matter of prudence or defense-in-depth would not be acceptable, given the great difficulties in conveying to the public that conformance with such objectives is not necessary to ensure adequate protection of the public health and safety and the environment or that conformance provides a substantial increase in protection beyond that necessary for adequate protection. Absent such a clear statement, it is likely that sites not meeting such general objectives would generate public concerns that the sites were unsafe for nuclear power plants.

(3) Defer consideration of any changes to the siting regulation until after the source term update and severe accident rulemakings are completed.

Specifically, complete source term update program, followed by establishment of severe accident requirements through rulemaking, followed by consideration of revisions to the siting regulations.

Postponement of fur,ther consideration of any revisions to the siting-rule -if the eommission-decides n0tc- -to-terminate the rulemaking proceeding would be acceptable. If the Commission chooses this route, the proposed rule should be withdrawn and the rulemaking proceeding terminated, at least with respect to the changes related to demography.

This is not meant to imply that ISG Members are in favor of revising Part 100 since deferral would not automatically result in an acceptable revision to Part 100. However, deferral would potentially ensure consistency among the source term update, the technical bases for any severe accident requirements, and management of re_sidual risk from a severe accident through the siting process. One of the problems with the proposed revisions is that they are intended to manage residual risk by elevating population over other siting factors, yet they do not reflect or take into account the residual risk from a severe accident.

The downside of postponement is that a revised Part 100 might not be avail&ble to applicants for an Early Site Permit under Subpart A to Part 52 .. However, this would in no way impede the safety evaluation of the site and ensuring that the site was suitable for a nuclear power plant. As the analysis in our original submission amply demonstrates, the present framework

N"EWMAN & HOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 9 works well in ensuring the selection of suitable nuclear power plant sites.

Issue No. 5: Extent to which reactor siting criteria conform to stated risk objectives, such as the Safety Goal, and the extent to which emphasis should be given to less quantifiable objectives such as defense-in-depth or prudence.

Because the present siting regime has worked well and because internationai safety practices reflect NRC s~fety practices to such a degree, it is inadvisable to change NRC's present siting regime unless there is risk reduction value from the changes which outweighs the costs -- including those to the international community. Basing changes on defense-in-depth or prudence arguments poses significant difficulties with respect.to both operating plants and advanced designs. The problem is that more restrictive requirements suggest that the risk is significant enough to warrant the changes, on the one hand, and that the proposed changes will significantly affect the risk, on the other. This is unlikely because of the nature and size of the risk. Also, regulators have historically had difficulty in allaying concerns once raised. The questions are asked: (1) If there is no basis for concern, why make the change? (2) If there is a basis for concern, shouldn't more be done?

In sum, unless there is a clear need to change the regulations, then the regulations should not be changed.

Issue No. 6: Appropriate balance between deterministic and probabilistic seismic evaluations.

In general, the Commission has moved carefully in the direction of risk-based regulation. However, the proposed revisions to the seismic requirements do not seem to reflect this cautious approach. The proposed revisions require both deterministic and probabilistic methods to be used in evaluating the suitability of sites, but do not provide reconciliation methods should application of the methods yield different results.

NEWMAN & HOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 10 Issue Ho. 7z Extent to which t ~ g of proposed revisions are being driven by the prospects of an early site perndt.

Changes to Part 100 should not.be driven by prospects of an early site permit. The Commission's present siting framework ensures the selection of suitable sites consistent with remote siting objectives. Nothing in the Commission's seismic requirements precludes application of probabilistic methods in the seismic evaluations. The impacts of changing the regulatory requirements are sufficiently great that change should be approached cautiously, not precipitously.

Issue Ho. 8z Extent to which proposed revisions support the Commission policy of consistent and predictable

• practice ( ~ , the issue of assurance versus, flexibility afforded by the proposed revisions).

Issue No. 8 assumes implicitly that it is not necessary to have flexibility in order to select sites which are among the best reasonably to be found. It suggests that diminished flexibility has no safety implications and, therefore, that more prescriptive siting requirements can be established to achieve consistent and predictable practice. In"other words, Issue No. 8 seems to imply that flexibility in selecting suitable sites and consistent, predictable practice can be traded off without affecting the quality of the sites selected.

ISG Members do not agree with the premise implicit in Issue No. 8. As discussed above under Issue No. 3, selecting sites among the best reasonably to be found involves consideration of 14 siting parameters. The weight to be assigned each of the 14 varies from site to site and region to region.

Flexibility is necessacy to ensure appropriate weighing in a manner which provides for adequate protection of the public health and safety and the environment. Elevating the importance of one, such as population density, over all of the others, including seismic and hydrologic characteristics of a site, potentially has safety implications that could lead to the selection of sites that are not among the best reasonably to be found.

Thus, the issue of tradeoffs between consistent, predictable practice and flexibility in the case of siting is not a valid issue. Moreover, as we discussed above, as long as the public has questions about the use of nuclear power and the

NEWMAN & HOLTZINOER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 11 licensing forum remains available to raise such questions, contentions will be raised concerning the location of nuclear power plants. Inappropriately restrictive siting requirements without a safety basis, such as the proposed changes to Part 100, will not limit the frequency with which contentions are raised in the adjudicatory context of licensing.

Issue Ho. 91 Plans to ensure that there is feedback between the source term development effort and the severe accident rulema.king process.

Ensuring feedback between the source term development effort and the severe accident rulemaking process is very important. Source terms selected as the basis for regulation should reflect the reactor designs being regulated. Likewise, severe accident requirements should be appropriate to the designs being regulated. In establishing,severe accident requirements, it will be important to understand the risk reduction potential of proposed requirements. To do this, it will be necessary to perform dose calculations using appropriate source terms. This has schedular implications; namely, the source term update should precede finalization of severe accident requirements.

Not only should there be feedback between source term development and the severe accident rulemaking, there should be feedback between these activities and any consideration of revisions to the siting requirements. As we discussed above under Issue No. 1, any revisions to the siting requirements should be based on risk reduction potential. To evaluate risk reduction potential, dose calculations must be performed. The TIO 14844 source terms are obsolete and should not be used in such calculations. A clear implication of this view is that further consideration of revisions to the siting regulations should await completion of the source term update and severe accident rulemaking proceeding.

The Commission also asked the Staff to consider the desirability of holding an international workshop or other suitable forum for interaction on these issues. We are pleased to comment on that question. We do not believe it is necessary to hold an international workshop in order to obtain the views of the international community on these issues. We believe that solicitation of views through the notice and comment process provides adequate and economical opportunity to obtain these views. Several opportunities should occur as the Commission gives further consideration to alternatives. For example, the

NBW'MAN & HOLTZINGER, P. C.

Mr. Samuel J. Chilk October 1, 1993 Page 12 Commission might offer opportunity to comment on the SECY paper transmitting the Staff's analysis of alternatives to the proposed revisions, which is being prepared in response to the Commission's August 12th SRM. Additionally, should the Commission proceed with proposal of alternative revisions to the siting regulations, another opportunity to solicit the views of the international community would be presented.

Very since

• William O.

L. Manning Janet E.B.

ccz Chairman Ivan Selin Commissioner Kenneth c. Rogers Commissioner Forrest J. Remick Commissioner E. Gail dePlanque Mr. William C. Parler, General Counsel Mr. James M. Taylor, Executive Director for Operations Dr. Eric s. Beckjord, Director Office of Nuclear Regulatory Research Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation Mr. Carlton Stoiber, Director Office of International Programs Dr. J. Ernest Wilkins, Jr., Chairman Advisory Committee on Reactor Safeguards Dr. Themis Speis, Deputy Director for ~esearch, Office of Nuclear *Regulatory Research Mr. Warren Minners, Director, Division of Safety Issue Resolution, Office of Nuclear Regulatory Research Mr. Charles Ader, Chief, Severe Accident Issues Branch, Office of Nuclear Regulatory Research Mr. Leonard Soffer, Office of Nuclear Regulatory Research

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~*Li Transmittal INTERNATIONAL l'I

- 1 JJJ-93-0330 July 12, 1993 Mr. Roger Kenneally USNRC NRR/MS-NL/5-Room 007 5650 Nicholson Lane Rockville, MD 20852

Subject:

Comments on "Appendix S to Part 50 -- Earthquake Engineering Criteria for Nuclear Power Plants."

Dear Roger:

Thank you for the opportunity to comment on the subject document.

My comment focuses on Sec. N Application to Engineering Design. In particular, the statement:

"At a minimum, the horizontal Safe Shutdown Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.lg."

US NRC Standard Review Plan (SRP) Sec. 2.5.2 states the free-field ground motion (also, referred to as control motion) should be defined to be on a ground surface and should be based on data obtained in the free-field. Two cases are identified: a relatively uniform soil or rock profile for which the control motion is defined at top of finished grade in the free-field; and sites with one or more thin soil layers overlying a competent material, where the control motion is specified on an outcrop or a hypothetical outcrop corresponding to the top of the competent material. The point at which the motion is specified is denoted the control point. In all cases, the free-field ground motion should be compatible with the properties of the soil profile.

US NRC SRP Sec. 3.7.2 requires the deconvolved motion at foundation level in the free-field to envelope 60% of the free-field surface motion. For soil sites, the envelope of the three soil property cases can be used to satisfy the criteria.

In my judgment, these two sets of criteria adequately address concerns regarding lack of frequency content or amplitude of ground motion for design. In fact, one could effectively argue that based on recorded free-field ground motion data, the 60% rule applied with no regard for depth of embedment may be too conservative.

44 Montgomery Street, Suite 3200 San Francisco, CA 94104 USA Telephone (415) 989-2000 FAX (415) 433-5107

Mr. R. Kenneally July 12, 1993 Page2 of2 Appendix S to Part 50 with regard to this issue, may be contradictory to the SRP and is non-specific in its present form.

1. What is an "appropriate response spectrum?" Does it include reductions at frequ encies corresponding to frequencies of the embedment layer?

2. The 0.1g PGA is likely met when enforcing the 60% rule. However, for soil sites, does the envelope of the three soil property cases apply?

New designs, such as ALWR, are effectively using embedment to mitigate earthquake effects on structures, systems, and components. The regulations should not penalize this approach.

In summary, this requirement should be re-evaluated in light of the SRP and the above comments. If it remains, a definitio n of the applicable terms and a specification of methods to satisfy the criteria should be included.

Please find enclosed a recent article on soil-structure interaction highlighting much recent data which has been acquired.

I look forward to discussing this issue with you in more detail and at your convenience.

Sincerely, EQE International, Inc.

9-*~~r Dr. James J. Johnson President/EEC (415)989-2000 JJJ:emb

Enclosure:

Johnson, J. J. and A. P. Asfura, "Soil -Structure Interaction (SSI):

Observations, Data, and Correlative Analysis," Proceedings from the NATO Advanced Study Institute on Developments in Soil-Structure Interaction, Ankara, Turkey, 1993 emb/jjjken

SOIL-STRUCTURE INTERACTION (SSI):

OBSERVATIONS, DATA, AND CORRELATIVE ANALYSIS JAMES J. JOHNSON ALFJANDRO P. ASFURA EQE International, Inc.

44 Montgomery Street, Suite 3200 San Francisco, CA 94104 Abstract Soil-structure mteracUon (S SD has been reVIewed in hght of ground mollOn and structmal. response data accumulated to date. Correl.attons of observatlons With site response and SSI anal yscs arc presented. The S SI analystS process is dlSCUSSCd accordmg to the component elements: specrlicatlon of the free-field ground monon, models of sotl and structures, SSI analys]S, and mteqntanoo of responses. Observed data and correlative analyses help clanfy the nnporumt LSSUeS mvolved ma SSI analysis, the quahty of the 118SUIIlptlons made m modem practtce, and therr impact on structure response.

1. Introduction 1.1. SSI: A STATEMENTOFTIIEPROBLEM Toe respoffie of a structure during an earthquake depends on the characteristics of the ground motion, the surrounding soil, and the structure itself. For structures founded on rock or very stiff soils, the A foundation motion is essentially that which would exist in the soil at the level of the foundation in the WI' absence of the structure and any excavation; this motion is denoted the free-field ground motion. For soft soils, the foundation motion differs from that in the free field due to the coupling of the soil and structure dunng the earthquake. lbis interaction results from the scattering of waves from the foundation and the radiation of energy from the structure due to structural vibrations. Because of these effects, the state of deformation (particle displacements, velocities, and accelerations) in the supporting soil is different from that in the free field. In turn, the dynamic response of a structure supported on soft soil may differ substantially In amplitude and frequency content from the response of an identical structure supported on a very stiff soil or rock. The coupled soil-structure system exhibits a peak structural response at a lower frequency than would an identical rigidly supported structure. Also, the amplitude of structural response is affected by the additional energy dissipation introduced into the system through radiation damping and matenal damping in the soil.

219 P. GuUcan and R. W Clough (t!ds ), IMvelopmen!s m Dynamic So1/-Str,,ctun lnkractwn, 219-258 0 1993 Khtwer AcmJmuc Pubbshen Prmud m 1M Netherlands.

220 The analysis of SSI depends on the specification of the free-field ground mot.ton and the Idealization of the soil and structure. Modeling the soil entails its configuration and material properties. Modeling the structure includes the 81.I'uCture itself and its foundation. '!"re calculated I'eS(XJnses must be interpreted ln light of differences between the idealized system and the real physical situation and the uncertainties known to exist In the phenomena. Figure 1 shows the elements of seislDlc analysis necessary to calculate seismic response including SSI effects. Table 1 lists the various aspects, including Interpretation and recognition of uncertainties In the process.

The state of knowledge of SSI was well documented in 1980 by Johnson [1]. Reference 1 contains solicited contributions from key researchers (Luco [2], Roesset [3], Seed and Lysme.r [4]) and drew upon other researchers and practitioners as well (Veletsos [5], Chopra [6]). Reference 1 provided a paradigm for discussions over the decade of the 1980s.

Toe ensuing decade was characterized most by the accumulation of substantial data supporting and clarifylng the roles of the various elements of the SSI phenomena. It is this data and its evaluation that is the primary focus of this paper. In additlon to data acquisition and evaluation, significant progress has been made in SSI analysis techniques which include the implementation of boundary element methods. Also, the 1980s has seen a revision of Important regulatory practice for nuclear power plants in the U.S. to conform to the current state of knowledge.

'!"re remainder of Sec. 1 describes the specification of the free-field ground motion and the idealization of the soil and structure, ln general terms, {r(}Vidlng the basis for subsequent discussions of observatiolllll data and its evaluation. Finally, a preview of the remaining sections and the introduction of the most comprehensive SSI experiment performed to date is presented.

1.2. SPECIFICATION OF THE F'REE-FJBI.l) GROUND MOTION Descnblng the free-field ground motion at a site for SSI analysis purposes entails specifying the point at which the motion is applied (the control point), the amplitude and frequency characteristics of the motion (referred to as the control motion and typically defined as grouoo response spectra, power specttal density functions, and/or time histories), the spatial variation of the mot.ton, and, in some cases, duration, magnitude, and other earthquake chllracteristics. In terms of SSI, the variation of monon over the depth and width of the foundation is relevant. For surface foundations, the variation of motion on the surface of the soil is important; for embedded foundations, the variation of motion ove.r both, the embedment depth and the foundation width. should be known.

1.3. MODEUNO Tiffi SOII..-STRUCTURE SYSTEM Modeling the soil-structure system entails modeling:

Soil Profile. Soll/rock layering ln the site vicinity and the dynamic soil/rock properties.

Soil Nonlinear Behavior. The variation of dynamic soil/rock ixoperties as a function of the strain induced by the seismic waves. Soll shear modulus and material damping vs. shear strain relationships define the variation at a point.

Foundation. Three aspects are important embedment, stiffness, and geometry.

Structurt!. Dynamic behavior, including linear and nonlinear aspects.

SS! Parameters and Analysts. Several ways of categorizing SSI analysis methods have been A used. Two ways are (1) direct methods which analyze the soil-structure system in a single step

• and (2) the substructure approach which treats the problem in a series of steps, e.g., dete.rmination of the foundation input motion and the foundation nnpedances, modeling of the structure, and the

221 analysis of the coupled system. In the context of this paper, it ls informative to view the SSI phenomenon in the steps of the substructure approach. Figure 1 showed schematically the substructure approach. 'The key elements not previously ~ are:

Foundation input motwn. Foundation input motion differs from the free-field ground motion in all cases, except for surface foundations subjected to vertically incident waves, provided the spatial variation of the free-field ground motion is taken into account. This is due, first, to the variation of free-field motion with depth in the soil and second, due to the scattering of waves from the soil-foundation interface because points on the foundation are constrained to move according to its geometry and stiffness.

Fowuiatwn impedances. Foundation impedances describe the force-displacement characteristics of the soil. They depend on the soil configuration and material behavior, the frequency of the excitation, and the geometry of the foundation. In general, for a linear elastic or viscoelastic material and a uniform or horizontally layered soil deposit, each element of the impedance matrix is complex-valued and frequency dependent For a rigid foundation, the impedance matrix ls 6x6, which relates a resultant set of forces and moments to the six rigid-body degrees-of-freedom.

Soil-structure interaction analysis. The final step in the substructure approach is the actual analysis. 1be result of the previous steps: foundation input motion; foundation impedances; and structure models are combined to solve the equations of motion for the coupled soil-structure system. The entire process ls sometimes referred to as a complete interaction analysis which is separated into two parts: kinematic interaction described by the scattering matrix; and inertial interaction comprised of the effects of vibration of the soil and structure.

~ terms of SSI parameters, the scattering functions model kinematic interaction and the 9Jundatlon impedances model inertial interaction. It ls these two parameters which are highlighted in the ensuing sections.

1.4. OR0ANl7ATION AND PERSPECTIVE This paper is organized according to the elements of SSI as depicted in Table 1. Each of these elements has been introduced previously and the ensuing sections focus on observational data and its analytical treatment Section 2 contains control motion, earthquake characteristics, control point location, and the combined topics of spatial variation of ground motion and kinematic interaction effects; important aspects of the specification of the free-field ground motion. Section 3 discusses soil modeling for site response and SSI analyses. Section 4 discusses the remaining aspects of SSI including foundation impedances, the response of soil-structure systems, and special issues such as structure-structure interaction. Section 5 presents conclusions and perspectives. This approach is similar to that of Johnson and Chang [7] and includes selected data from their work.

Before proceeding, an important SSI experiment is discusred. Very few opportunities exist to record free-field motion and structure response for an earthquake. In the mid 1980s, the Electric Power Research Institute (EPRI), in cooperation with Taiwan Power Co. (TPC), constructed two-scale model reinforced concrete containment buildings (1/4 and 1/U scale). Toe experiment was extensively instrumented in the free-field and in the structure. It was located within a strong-motion array (SMART-1, Strong Motion Array Taiwan, Number 1) sponsored by the US National Science Foundation and maintained by the Institute of Earth Sciences of Academia Slnlca of Taiwan. The goal of the experiment was to measure the responses at instrumented locations due to vibration tests and actual earthquakes.

222 Using this data base, a cooperative program to validate SSI analysis methodologies was sponsored by EPRI, 1PC, and the US Nuclear Regulatory Commission (NRC). Numerous publications document the results of the SSI analysis studies. A two-volume EPRI report [8] contains the proceedings of a workshop held in Palo Alto, CA, USA on December 11-13, 1987 to discuss the experiment, data collected, and analyses perfonned to investigate SSI analysis methodologies and their application. Johnson et al [9] performed one set of analyses from which results are presented here to demonstrate various aspects of the SSI phenomenon. Recently, a summary of lessons learned was published [10].

When appropriate throughout this paper, data from the Lotung experiment is presented and discussed.

2. Speclftcation of the Free-field Ground Motion Specification of the free-field ground motion for SSI analysis of structures entails specifying the control motion, the control point, the spatial variation of the motion, and descriptors of the earthquake such as magnitude, duration, location of the earthquake relative to the site, etc. The control motion and the descriptors of the earthquake are discussed next Control point location, spatial variation of the free-field ground motion, and selected aspects of kinematic interaction are discussed in a later subsection.

2.1. CONTROL MOTION AND EARTiiQUAKE CHARACTERISTICS The control motion is defined by specifying the amplitude and frequency content of the earthquake to be considered. Two purposes exist for SSI analysis of a structure: design or evaluation of a facility for a specified earthquake level; or evaluation of a structure for a specific event. In the funner case, statistical combinations of recorded or predicted earthquake motions are typically the bases. In the latter case, recorded earthquake acceleration time histories typically comprise the free-field ground motions.

The frequency content of the motion is one of the most important aspects of the free-field motion as it affects structure response. For linear structural behavior and equivalent linear SSL the frequency content of the free-field motion compared to important frequencies of the soil-structnre system determines response. For structures expected to behave inelastically during the earthquake (and, in particular, structures for which SSI is not important), structure response is detennined by the frequency content of the free-field motion, i.e., in the frequency range from the elastic frequency to a lower-frequency corresponding to a certain amount of inelastic deformation.

2.1.1. Aggregated Ground Motions. A wide variety of ground response spectra have been specified for design of major facilities, such as nuclear power plants, depending on the vintage of the plant and the site soil conditions. The majority of these have been relatively broad-banded spectra representing a combination of earthquakes of different magnitudes and distances from the site. Conmruction of -

such design spectra are usually based on a statistical analysis of recorded motions and frequently targeted to a 50% or S4% nonexceedance probability (NEP). Three pomts are important relative to these broad-banded spectra. FI.I'st, earthquakes of different magnitudes and distances control different frequency ranges of the spectra. Small magnitude earthquakes contribute more to the high frequency

223 range than to the low frequency range and so forth. Second, it Is unlikely that a single earthquake will have frequency content matching the design ground response spectra. Hence, a degree of conservatism Is added when a broad-banded response spectra defines the control motion. Third, a single earthquake can have frequency content that exceeds the design ground response spectra in selected frequency ranges. Toe likelihood of the elfceedanre depends on the NEP of the design spectra. Figure 2 compares several ground response spectra used in the design or evaluation process.

U.S. NRC Regulatory Guide 1.60 [11] response spectra defined design criteria for nuclear power plants designed after about 1973. These spectra were targeted to an 84% NEP of the data considered but exceed this target in selected frequency ranges. Figure 2 shows an additional set of broad-banded spectra for rock and alluvial sites and for 50% and 84% NEP. U.S. NRC NUREG/CR-0098 [12] is the source. These spectra are currently being used extensively for defining a seismic margin earthquake for which seismic margin assessments are being performed for U.S. nuclear power plants.

Also, the broad-banded spectral shape is being used to define design cnterla for new design.

In addition to the site independent ground response spectra discussed above, two additional forms of ground response spectra are being generated and used for site specific design or evaluations. First, site specific spectra are generated by accumulating recorded data that meets the design earthquake characteristics and local site conditions, analyzing the data statistically, calculating ground response spectra of various statistical attributes, a¢ selecting response spectra for design or reevaluation.

Figure 3 shows an example for a specific site; the 84% NEP was selected. Second, seismic hazard studies are performed to generate farnilles of seismic hazard curves which yield estimates of the probability of exceedance of earthquakes with specified peak ground acceleration (PGA) values or

~ - Confidence limits for these seismic hazard estimates are derived from the family of curves.

~ompanion to the seismic hazard curves are uniform hazard spectra (UHS) which are ground response spectra geIWllted for a specified renun period or probability of exceedance level and with various confidence limits. Figure 4 shows example UHS for a specific site, 10,000 year return period, and 15%, 50%, and 85% confidence limits. Such spectra are generated by the same probabilistic seismic hazard methodology as is used in generating the seismic hazard curves for PGAs.

In almost all cases, design ground response spectra are accompanied by ground motion acceleration time histories - artificial acceleration time histories generated with response spectra that match or exceed the design ground response spectra. Due to the enveloping process, additional conservatism is introduced. Ground motion time histories are used to generate In-structure response - loads, response spectra, displacements, etc.

2.1.2. Individual Recorded Events. The previous sections discussed aggregated motions as derived from recorded data or empirical models based on recorded data. 1bese aggregated motions do not represent a single event 'They have been developed for various design and evaluation purposes. It is informative to present recorded motions from actual earthquakes and visualize differences between a single event and the aggregated motions. Two earthquakes of note from an SSI standpoint were the May 20, 1986 and November 14, 1986 events which affected the Lotung scale model structures.

Figure 5 contains response spectra generated from the acceleration time histories recorded on the soil-free surface for the May 20, 1986 event. In the literature, these two events are denoted numbers 7 and 16, respectively. Each earthquake produced PGAs greater than about 02g in the hodzontal direction with principally low frequency motion, i.e., less than about 5 Hz. Subsequent sections will discuss these and other motions recorded in this area.

224 2.1.3. Magnitude E,jfects. Ground motion frequency content ls strongly dependent on specific factors of the earthquake and site. Two particularly important characteristics of the earthquake are its magnitude and epicentrnl distance from the site. Small magnitude events are characterized by narrow-banded response spectra and high frequency in comparison to moderate magnitude events.

Figure 6 [13] illustrates the effect of magnitude on response spectral shape. In the figure, the response spectral shape obtained from a series of small magnitude (ML< 4) earthquakes is compared with the response spectral shape from a moderate magnitude (ML "" 6 112) event. As shown, the small magnitude earthquakes are characterized by a narrow-banded response spectral shape and greater high frequency content than the moderate magnitude event 2.1.4. Soil vs. Rock Sites. One of the most important parameters governing amplitude and frequency content of free-field ground motions is the local site conditions, i.e., soil vs. rock: and shallow, soft soil overlying a stiff soil or rock. Two sets of data demonstrate dramatically the difference in motions recorded on rock vs. soil fur the same earthquake and sites in close proximity.

Toe Ashigara Valley in Japan is located about 80 km southwest of Tokyo. A digital strong-motion accelerograph array network [14] was installed Jn this seismically very active area. The geological profile of the Ashlgara Valley is shown in Fig. 7 [24]. The valley is an alluvial basin with rock outcrops at the mountain side and soft sedimentary soil layers at the basin. Accelerographs were illStll.lled in the rock outcrop (KR.1 ), at the surface of the soft layers (KSl and K2S) and inside the soil (KD l and KD2). Figures 8 and 9 [ 15] show response spectra of motions recorded at the rock outcrop (KRl) and at the surface of soft layers (K2S) for an earthquake occurring on August 5, 1990. These figures clearly show differences in motions due to different site conditions. High frequencies about 10 Hz and greater are dominant at the rock site and frequencies below about 5 Hz are d o m i ~

oa at the soil surface. Figures 8 and 9 clearly show the large amplificatlon of the low frequency waves due to the soil response and the deamplification of the high frequency waves due to the filtering effect of the soft soil.

The Loma Prieta earthquake, October 17, 1989, with epicenter near San Francisco, CA, provided the opportunity to collect and evaluare recorded motions on rock and soft soil sites, in some cases immediately adjacent to each other. Figures 10 and 11 [15] compare response spectra fur recorded motions on rock (Yerba Buena Island) and on soil (downtown Oakland). Significant differences in amplification are obvious with the rock motions being less. Figure 12 [16] presents amplification factors for soil (Tre$ure Island) vs. rock (Yerba Buena Island). For horirontal motions, significant amplification of soil over rock is apparent For vertical motions, rock values are higher than soil which requires additional evaluation. Note, Treasure Island and Yerba Buena Island are the north and south ends of a single land mass. Figures 13 and 14 (16] compare amplificatlon factors fur a number of soft soil sites with expected rock or stiff soll motions corrected fur distance according to appropriate attenuation laws. Significant amplification for soft soil vs. stiff soil or rock is apparent for horizontal and vertical motions except for the Treasure Island vertical component, discussed above.

The evidence supporting the differences between rock and soft soil motions continues to mount emphasizing the effect of local site conditions on the amplitude and frequency content of the motion.

2.2. CONTROLPoINT The term control point designates the locatlon at which the control motion is defined. The control point should always be defined on the free surface of soll or rock: at the site of interest. Specifying

225 the control point at locations other than a free surface ignores the physics of the problem and the source of data used in defining the control motion. Past regulatory practice has specified the control point to be at foundation level In the free-field, which is technically untenable. It not only ignores the physics of the problem and the source of recorded data, it also results in motions on the free surface whose response spectra display peaks and valleys associated with frequencies of the embedment layer which are totally fictitious and umealistic. The frequencies of these peaks correspond to the frequencies of the soil layer between the foundation and the free surface. These free surface motions are dependent on the foundation depth rather than free-field site characteristics. In addition, the peak acceleration of the resulting free surface motion is typically calculated to be significantly greater than the control motion. Hence, by all seismological definitions, a design or evaluation earthquake is increased. A 0.25g earthquake may become a 0.35g earthquake or greater depending on the embedment depth, soil properties, and control motion characteristics.* Fmally, specification of the control point at foundation level rather than the soil free surface effectively penalizes partially embedded structures compared to surface-founded structures which contradicts common sense and observations.

Simplistic SSI analyses frequently Ignore kinematic interaction for embedded foundations. In so doing, the foundation Input motion is assumed to be identical to the control motion. Implicit in this assumption is the definition of the control point as any point on the foundation and no spatial variation of ground motion. 1his assumption is almost always conservative and frequently extremely conservative. Recognition of this conservatism is important in interpreting the results of the SSI analysis.

~- SPA11AL VARIATION OF MOTION AND KINEMATIC INTERACTION Spatial variations of ground motion refer to differences in amplitude and/or phase of ground motions with horizontal distance or depth in the free-field. Spatial variations of ground motion are associated with different types of seismic waves and various wave propagation phenomena including reflection at the free surface, reflection and refraction at interfaces and boundaries between geological strata having different properties, and diffraction and scattering induced by nonuniform subsurface geological strata and topographic effects along the propagation path of the seismic waves. A vertically incident body-wave propagating in such a medium will include ground motions having identical amplitudes and phase at diffe{ent points on a horizontal plane (neglecting source-to-site attenuation effects over short horizontal distances). A plane wave propagating horizontally at some apparent phase velocity wiII induce ground motion having identical amplitudes but with a shift in phase in the horizontal direction associated with the apparent horizontal propagation velocity of the wave. In either of these ideal cases, the ground motions are considered to be coherent in that amplitudes (acceleration time histories and their response spectra) do not vary with location in a horizontal plane. Incoherence of ground motion, on the other hand, may result from wave scattering due to inhomogeneities of soil/rock: media and topographic effects along the propagation path of the seismic waves.

In terms of the SSI phenomenon, variations of the ground motion over the depth and width of the foundation are the important aspects [l]. For surface foundations, the variation of motion on the surface of the soil is important; for embedded foundations, tbe variation of motion on both the embedded depth and foundation width is Important Spatial variations of ground motions are discussed in terms of variations with depth and horizontal distances.

226 2.3.1. Variation of Ground Motion with Depth of Soll. For either vertically or non-verttcally incident waves, ground motions vary with depth. 11iese variations can gerexally be expressed in tenns of peak amplitudes. frequency content, and phase. Variations of ground monon with depth due to vertically and non-vertically incident body waves and surface waves have been extensively studied analytically by many investigators, Chang et al [13). These studJ.es, based on the physics of plane wave propagation tlrrough layered media, all indicate that, due pdmarily to the free-surface effect, ground motions generally decrease with depth. The nature of the variation is a function of frequency content and wave type of the incident waves, soil layering, and dynamic soil properties of each soil layer, including shear and compressional wave velocities, damping ratio, and lllMS density.

Free-field Motion. A review and summary of observational data on the variations of earthquake ground motion with depth was conducted by Seed and Lysmer [4] and Chang et al [13) reflecting the state of knowledge as of 1980 and 1985, respectively. Recordings from two downhole arrays in Japan were analyzed [13] and a review of published data from a number of downhole arrays in Japan and the U.S. was conducted. Based on the review of these data on the variations of ground motion with depth, it was concluded: there is a good body of a data to show that, in general, both peak acceleranon and response spectra decrease significantly with depth in the range of typical embedment depths of structures, i.e., less than approximately 25m; and it appears that deconvolution procedures assuming vertically propagating shear waves provide reasonable estimates of the variations of ground motion with depth.

Table 2 summarizes the sources of data evaluated by Seed and Lysmer [4] and Chang et al [13].

Since 1985, subsrantial additional data has been recorded and evaluated [20-26]. These sources arA included in Table 2 and four of the most relevant are discussed next

• Lotung, Taiwan. As part of the Lotung experiment, downhole free-field data was recorded at depths of 6m, 1 lm, 17m, and 47m. Figure 15 [8] shows the configuration of the arrays. Extensive studies investigating analytical modeling of the phenomenon have been performed [9, 20, 21, 26]. Figure 16 compares recorded and analytically predicted response for one of the models. For this case and many others, deconvolution was applied with the control motion being the free.surface recorded motion.

Equivalent linear soil properties were developed through an iteration process and used Other studies have investigated nonlinear soil behavior [20) to gain further insight into its importance. This set of recorded data and analytical predictions clearly support the variation of motion with depth in the soil and, in fact, a reduction of motion with depth as expected. In addition, the a&Wmption of vertically propagating waves and an equivalent linear representation of soil material behavior well models the wave propagation phenomenon especially at depths in the soil important to SSL Nonlinear material behavior was shown to have an effect when starting from recorded motion at depth (17m) and convolving the motion to the surface. Figure 17 [20] demonstrates this effect Turkey F1at, Ca. Cramer [22] and the compilation of Ref. 23 report on data recorded at Turkey F1at, Ca Although the sources of motion were low magnitude, a comparison of the surface and downhole data show clear reductions of peak values with depth in the soil Turkey F1at includes instruments in rock and soil with consistent results. -

Ashigara Valley. An excellent collection of papers is presented in Ref. 23 regarding recorded and predicted responses in the Ashigara Valley network. Figure 7 shows a profile of the site and the location of the instruments. Figure 18 shows peak ground acceleration data recorded and predicted at

'227 two surface and one downhole location. Again, the reduction of motion with depth can be clearly seen. Figure 19 compares response spectra recorded and predicted at the surface and at depth In the soil. Midorikawa [24] gives a summary of the studies performed for the Ashigara Valley.

Garnes Valley, Ca. Seale and Archuleta [25] present data recorded through the end of April 1990 in the Games Valley surface and downhole array located at the San Jacinto fault zone. They present individual data from several events that demonstrate the variation of motion with depth of soil.

Reductions m amplitude are apparent Foundation Response. A comparison of motions recorded in the free-field and on the base of partially embedded structures provides excellent data validating the effects of kinematic Interaction and the spatial variation of ground motion. All recorded data on structnres Includes the effects of SSI to some extent Section 1.3 introduced the concept of kinematic and inertial interaction. For purposes of validating the spatial variation of motion with depth In the soil, observations of kinematic interru.,'tion are sought Toe ideal situation Is one where free-field surface motions and foundation motions are recorded for a structure whose embedded portion is stiff, approaching rigid behavior, and the dynamic characteristics of the structure are such that inertial interaction is a minimal effect Two types of data are available.

Free-field surface motions vs. foundation response. Typically, differences in peak values, time histories, and response spectra were observed and transfer functions relating foundation response to

-ee-field surface motion were generated. These frequency-dependent transfer functions are, in

~

• one element of the scattering matrix when inertial Interaction effects are minimal, i.e., the component relating foundation input motion to horizontal free-field surface motion. In no cases, was enough Information available to generate the rocking component of the scattering matrix from recorded motiom. To do so requires recorded rotational acceleration time histories or the ability to generate them from other measurements.

LNG tanks, Japan. Eighteen LNG tanks of varying dimensions and deeply embedded were instrumented and motions recorded on their foundations and in the free-field for a large number of earthquakes. Many of the earthquakes were microtremors but detailed response for at least one larger event was presented. Transfer functions between free-field surface motions and foundation response were generated and compared with a calculated ~ttedng element Results compared well. A significant reduction in foundation response from the free surface values was observed. Toe mass and stiffness of the tanks was such that kinematic interaction dominated the soil-structure interaction effects. Flgure 20 presents the data [27].

Humboldt Bay Nuclear Power Plant 1be Humbolch Bay Nuclear Power Plant has experienced numerous earthquakes over the years. Four of note are the Ferndale earthquake of June 7, 1975 and the recent Lost Coast earthquake sequence of April 25, 1992 [28]. Accelerometers in place in the free-field and on the base of a deeply embedded caisson structure (-80 ft.) recorded acceleration time histories. Figure 21 [4] compares free-field surface response spectra with those recorded on the

- caisson's base for the 1975 event A comparison of peak acceleratiom shows:

228 Peak Accelerations Surface/Caisson Earthquake E-W N-S Vertical 6rln5 0.35/0.16g 0.26/0.12g 0.06/0.lOg 4fl5/92 (11 :06 PD'I) 0.22/0.14g 0.22/0.llg 0.05/0.08g 4126/92 (00: 14 PD'I) 0.25/0.12g 0.23/0.Ug 0.05/0.12g 4126/92 (04:18 PD'I) 0.13/0.0?g 0.098/0.057g 0.031/0.037g For horizontal motions, significant reductions are apparent, i.e., reductions up to 55%. For vertical motions, a slight Increase ls observed which is inconsistent with other data and requires further Investigation. The dominant SSI pheoomenon here, as for the LNG tanks, is likely to be kinematic interaction due to the deep embedment of the caisson.

Hollywood Storage Building, California. Toe Hollywood storage building, a 14-story reinforced concrete structure with basement embedded 9 ft. and supported on piles, was subjected to the Kem County earthquake of 1952 and the San Fernando earthquake of 1971. Recorded motions in the barement and an adjoining parking lot permit a comparison of free-field surface motion ~

foundation motion. Reduced motion on the foundation Is observed (Fig. 22).

• Fukushima Nuclear Power Station. Toe Fukushima Nuclear Power Station was subjected to the Miyagi-Ken-Oki, Japan earthquake of June 12, 1978. Motions were recorded in the free-field and in Units 1 and 6 reactor buildings. A comparison of peak surface free-field horizontal accelerations with those on the foundation showed 0.10g vs. 0.08g in the north-south direction and 0.13g vs. 0.06g in the east-west direction. Unfortunately, time histories or response spectra are not generally available in the open literature to see the variations with frequency. One such comparison is available which shows a reduction in spectral acceleration over the entire frequency range, i.e., foundation response ls less than free-field values. Both kinematic and inertial interaction effects are likely to be Important for this case. ,

Pleasant Valley Pumping Station. The Pleasant Valley Pumping Station on the California aqueduct was subjected to the Coalinga, CA earthquake sequence of May 1983. Toe structure is embedded 22 ft. below plant grade and is instrumented at three locations. A comparison of foundation response with a fourth instrument considered to be free-field shows a significant reduction in motion. A complication in this case, however, is that this free-field instrument is located in an instrument shelter on top of a slope 70 ft. above plant grade. Consequently, topographical effects are likely to have influenced this free-field record. A direct comparison of It with the foundation motion demonstrates the combined effects of spatial variation of motion and scattering. Figure 23 shows a cross-section through the aqueduct and pumping station and instrument locations. Figure 24 compares surface free-field motions with foundations response for one horizontal direction.

Buildings With and Without Basements. A series of buildings in close proximity to each other, with and without basements, subjected to the San Fernando earthquake of 1971 were investigated.

229 Comparing recorded basement motions for buildings with and without embedment documents the effects of kinematic interaction. 1be results show a definite reduction in motion of embedded foundations compared to surface foundations or free-field surface values. Numerous comparisons are presented by Chang et al. [13]; one is shown in Figure 25 for illustration.

Model Structures. Two model structures in Japan and the Lotung scale model structures in Taiwan have been instrumented and have recorded ground motions from a number of earthquakes. One model structure is located at the Fukushima Nuclear Powe.r Plant site. Tanaka et al. [19] report on data recorded in the free-field and structure. A reduction in peak acceleration of about 20% ls observed from free-field surface motion to foundation [19]. The second model structure resides at Chiba Field Station, Institute of Industrial Science, University of Tokyo. The purpose of the model was to instrument it for earthquakes, record earthquake motions, and investigate variability in ground motion and response. Figure 26 shows a cross-section [29,30] through the scale model structure including components (piping, hanged tank, and vessel) and instrument locations. As of 1978, Shibata [29] shows graphically that the mean response factor for peak acceleration on the foundation compared to the free-field ls about 0.68 with a coefficient of variation of 0.125 in the horizontal direction and 0.83 with a coefficient of variation 0.136 in the vertical direction. A clear reduction in response of the foundation is observed. The third model structure is the Lotung, Taiwan case.

Comparing the May 20, 1986 measured free-field surface response with foundation response demonstrates the reduction in free-field motion with depth. Figure 27 shows a schematic of the Lotung one-quarter scale model with instrument locations identified. Figure 28 [9] shows response

~ of motions recorded on the foundation (F4LS) to be compared with F1gure 5.

2.3.2 Variation of Ground Motion in a Horir.ontal Plane. Variations observed in the motions of two points located on a horirontal plane are mainly due to the difference in arrival times of the seismic waves at the two points and to the amplitude and frequency modification that those seismic waves undergo due to the geotechnical characteristics of the media between the two points. With the ill5tallation of several dense accelerograph arrays (e.g., SMART 1 and LSST arrays in Taiwan, El Centro and Parldleld in california, and Chusal differential array in the former USSR) and with the recorded data collected from them, the spatial variation of ground motions have been quantified by defining a complex coherency function [31-38]. The coherency function of two motions is normally defined as their cross-power spectrum divided by the square root of the product of the powe.r spectra of the individual motions. As shown in Figs. 29 [31] and 30 [34], this function strongly depends on the distance between recording locations and on the frequency of the seismic waves. This spatial variation of ground motions is a critical element in the seismic analysis of structures on large foundations and in the analysis of long structures on multiple foundations. For these cases, coherency functions can be used to develop average motions for large foundations or to develop the proper motions (and their cross-correlation) at each support for long multiply-supported structures.

These data support a "base averaging" effect on free-field ground motions for large stiff foundations. In the absence of performing detailed SSI analyses Incorporating the coherency functions, a simpler approach may be taken, i.e., a filtering of motions. For a 50m plan dimension foundation, reductions in spectral accelerations of 20% in the greater than 20 Hz frequency range, of

- 10% to 15% in the 10 to 20 Hz range, and of5% in the 5 to 10 Hz range are supported by the data

230

3. Soil Modeling For soil sites describing the soil configuration (layering or straUg,.apb.y) and the dynamic material properties of soil are necessary to perform SSI analysis and predict soil-structure response.

Determining soil properties to be used in the SSI analysis is the second most unce.rtain element of the process; the first being specifying the ground motion. Modeling the soil can be visualized in two stages: determining the low strain in-situ soil profile, and associated mate.rial properties; and defining the dynamic material behavior of the soil as a function of the induced strains from the earthquake and soil-structure response. In general, dynamic stress-strain behavior of soils is nonlinear, anistropic, elasto-plastlc, and loading path dependent. It is, also, dependent on previous loading states and the degree of disturbance to be expected durlng construction. Practically speaking, these effects are not quantifiable in the current state-of-the-art and, hence, contribute to uncertainty in describing soil stress-strain behavior. Soil is modeled as a linear or equivalent linear viscoelastic medium in SSI analyses for earthquake motion, except in a few instances where nonlinear analysis has been performed for research objectives. A linear viscoelastic material model is defined by three parameters - two elastic constants (frequently shear modulus and Poisson's ratio although shear modulus and bulk or constrained modulus may be more appropriate) and material damping. Toe equivalent linear method 0.p{I"Oximates the nonlinear stress-strain relation with a secant modulus and material damping values selected to be compatible with the average shear strain induced during the motion using an iterative procedure. Requirements for this model are low strain shear modulus and Poisson's ratio (or bulk or constrained modulus), mareriaI damping values, and their variations with strain. Toe following discussion assumes an equivalent linear viscoelastic material model for soil A Three aspects of developing soil models are: field exploration, laboratory tests, and correlation 019' laboratory and field data.

Field Exploration. Field exploration, typically, relics heavily on boring programs which provide information on the spatial distribution of soil (horizontally and with depth) and produce samples for laboratory analysis. In addition, some dynamic properties are measured in situ, for example, shear wave velocity which leads to a value of sn:ar modulus at low strains.

Reference 39 provides a summary of field exploration, in general, and of boring, sampling, and in-situ testing, in particular. Low strain shear and compressional wave velocities are typically measured in the field. Various field techniques for measuring in situ shear and coffilYeSSional wave velocities exist including the seismic refraction survey, seismic cross-hole survey, seismic down-hole or up-hole survey, and surface wave techniques. The advantages and disadvantages of these techniques are discussed in Ref. 39. For sites having a relatively uniform soil pro:flle, all of these techniques are appropriate. However, for sites having layers with large velocity contrasts, the most appropriate technique is the cross-hole technique. Toe cross-hole data are expected to be the most reliable because of better control over and knowledge of the wave path. Toe downhole technique does not permit as precise resolution because travel times in any layer are averaged over the layer. Most field seismic survey techniques are capable of producing ground motions in the small shearing strain amplitude range (less than 10-3 percent). Hence, field exploration methods yield the soil IXOfile and estimates of important low strain materlal properties (shear modulus, Poisson's ratio, water table location).

Laboratory Tests. Laboratory tests are principally used to measure dynamic soil {I'Operties and their A variation with strain: soil shear modulus and material damping. Currently available laboratory

• testing techniques have been discussed and summarized in Ref. 39. These techniques include resonant column tests, cyclic triaxial tests, cyclic simple shear tests, and cyclic torsional shear tests.

231 Each test is applicable to diffe.rent strain ranges. 1be resonant and torsional shear column tests are capable of measuring dynamic soil properties over a wide range of shear strain (from 10-4 to 10-2 percent or higher). The c~lic triaxial test allows measurement of Young's modulus and damping at large strain (larger than Hr percent).

Typical variations of shear modulus with shear strain for clays, compiled from laboratory test data arc depicted in Flg. 31 [40]. Generally, the modulus reduction curves for gravelly soils and sands are similar. Flgure 32 shows material damping 88 a function of shear strain, also for clays. Shear modulus ~ and material damping increases with increasing shear strain levels.

Correlation of Laboratory and Field Data. Once shear modulus degradation curves and material damping versus strain curves have been obtained, h is necessary to correlate Iaboratocy detemrined low strain shear moclulus values with those in situ. Laboratory-meamired values of shear moduli at low levels of strain are typically smaller than those measured in the field. Several factors have been found to contribute to lower moduli measured in the laboratory. These factors include effects of sampling clistnrbance, stress history, and time (aging or period under sustained load). Thus, when the laboratory data are used to estimate in situ shear moduli in the field, considerations should be given to these effects. When laboratory data do not extrapolate back to the field data at small strains, one of the approaches used in practice is to scale the laboratory data up to the field data at small strains.

This can be done by proportionally increasing all values of laboratory moduli to match the field data at low strain values or by other scaling procedures.

Equipment LiMar Soil Properties. Given the low strain soil profile determined from the combination of field and laboratory tests, and the variation of material parameters with strain level, a site response Amaiys1s is performed to estimate equivalent linear soil properties for the SSI analysis.

W:.Tncertainty in Modeling Soil/Rocle at the Site. There is uncertainty in each aspect of defining and modeling the site soil conditions for SSI analysis purposes.

The soil configuration (layer or stratigraphy) is established from the boring program. Even after such a program, some uncertainty exists in the definition of the soil profile. Soils arc seldom homogeneous, and they seldom lie in clearly defined horizontal layers - the common assumptions in SSI analysis. In general, complicated soil systems Introduce a strong frequency dependence in the site behavior.

Modeling the dynamic stress-strain behavior of the soil is uncertain in two respects. First, modeling soil as a viscoelastic material with parameters selected as a function of average strains is an approximation to hs complex behavior. Second, there is uncertainty in defining low strain shear moduli and in defining the variations in shear modulns and observed damping with strain levels.

Variability can be observed in reporting test values for a given site and for reported generic test results [40,41,42]. Toe range of coefficients of variation on "stress-strain behavior estimated in Ref.

43 is 05 to 1.0. Considering only uncertainty in soil shear modulus itself and the eqnivalent linear estimates to be used in the SSI analysis, a coefficient of variation of 0.35 to 05 on soil shear modulus Is estimated. Uncertainty in low strain values is less than uncertainty in values at higher strains.

The Lotung SSI experiment provides a unique opportunity to quantify the uncertainty in estimates of equivalent linear soil properties. Toe bases for determining the equivalent linear soil properties included field and laboratory tests and the results of furcecl vibration tests on the stroctnre. The latter A responses permitted system identification techniques to be employed to refine estimates of the low

• strain profile.

Soil properties data was extracted from Ref. 8 and plotted in Figs. 33 and 34. Ten independent sets of data were available .and their variabfllty is apparent from the figures. To qnantlfy this variability, the data was evaluated statistically and a weighted COY calculated for soil shear modulus and

232 damping; the weighting factors WQ'C soil thickness to a depth of about 100 feet. Tue resulting COVs for shear modulus and damping were 0.48 and 039, respectively.

Hence, even for a most ideal situation. significant variability in soil properties as used in the SSI analysis should be expected reflecting variability innate in the soil, its behavior, and the modeling thereof. These estimates are for a single earthquake, the May 20, 1986 event.

4. Soll-structure Interaction Analysis 4.1 FoUNDATION MODELING Three aspects of modeling structure foundations are important: stiffness, embedment, and geometry.

Stiffness. The stiffness of a structural foundation is of importance because in almost all SSI analyses, by either direct or substructure methods, foundation stiffness is approximated. Most substructure analysis approaches assume the effective foundation stiffness to be rigid. For direct methods, various representations of foundation stiffness have been used depending on the geometry and other aspects of the structure-foundation system.

Most foundations of the type common to major bulleting structures canoot be considered rigid by themselves. However, structural load resisting systems, such as shear wall systems, /:igniflcantly stiffen their foundations. Hence, in many instances, the effective sti:ffress of'the foundation is verA high and it may be ~ rigid. One study which has investigated the effects of foundatlo-flexibillty on structure response for a complicated nuclear power plant structure of large plan dimension was performed [44]. In this study, the stiffening effect of the structure on the foundation was treated exactly. Even though the structure had nominal plan dimensions of 350 feet by 450 feet, the effect of foundation flexibility on structure response was minimal. The largest effect was on rotational accelerations of the foundation segment as- one would expect. Translational accelerations and response spectra, quantities of more interest, on the foundation and at points in the structure were minimally affected (5%). This study emphasized the importance of considering the stiffening effect of walls and floors on the foundation when evaluating its effective stiffness.

Hence, practically speaking, for design and evaluation of major structures with stiff load resisting systems, the asmnnptlon of rigid foundation behavior is justified. For predicting the response of a structure to a single recorded event, more refined modeling of the foundation may be required.

E:mbedment. Foundations are typically fully embedded, and structnres are partially embedded.

Foundation embedment has a significant effect on SSI. Both the foundation input motion and the foundation impedances differ for an embedded foumation compared to a surface foundation.

Foundation input motion for embe<kled foundations was discussed extensively in Sec. 2. Foundation impedances comprise a second aspect of foundation modeling. As discussed above, a common and appropriate assumption for many cases is rigid foundation behavior. For a rigid foundation, the impedance matrix describing force-displacement characteristics is, at most, a 6X6, complex valued, A frequency dependent matrix. The literature [1] contains numerous exact and !lp{XOximate analytical

• representations of foundation impedances. Also, many analytical/experimental correlations exist with relatively good results reported.

233 Figure 35 [45] shows one such example. Figure 36 [46] demonruates the effect of embedment on foundation response for embedded and nonembedded configurations of an assumed rigid foundation.

Forced vibration tests were performed. 'The results clearly demonstrate the effect on ioertlal interaction of embedment.

Geometry. The geometry of major structure foundations can be extremely complicated. Fortunately, many aspects of modeling the foundation geometry are believed to have second order effects on structure response; in particular, modeling the precise shape of the foundation in detail. However, other overall aspects such as nonsymmetry which lead to coupling of horizontal translation and torsion, and vertical translation and rocking, can be very important and need to be considered.

Modeling the foundation Is Important wi,th respect to structure response. This Is one area where knowledge and experience of the practitioner is Invaluable. Complicated foundations must be modeled properly to calculate best estimates of response, i.e., including the important aspects as necessary. Foundation modeling plays a key role in assesfilng whether two-<limenslonal models are appropriate or three-dimension models are required.

4.2 STRUCTURE MODELING 42.1. Linear Dynamic Bt!havior. Structures for which SSI analysis is to be performed require mathemancal models to represent therr dynamic characteristics. The required detail of the model is dependent on the complexity of the structure or component being modeled and the end result of the a,naiysis. Structure models are typically of two types: lumped-mass, stick models; and finite element

~odels. Lumped-mass, stick models are characterized by: lumped masses defining dynamic degrees-of-freedom, usually at floor slab elevations; simplified assumptions as to diaphragm or floor behavior are made, in particular floors are frequently assumed to behave rigidly, although diaphragm flexibility can be modeled If necessary; connections between lumped masses are usually stiffness elements whose stiffness values represent groups of walls running between the floor slabs; and, in some instances, an offset is modeled between the center of mass and center of rigidity at each floor elevation to account for coupling between horizontal translations and torsion. Finite element models are used to represent complex structures. They permit a more accurate representation of complicated situations without requiring significant simplifying assumptions. Either model type may be adequate depending on the structural con.figuration, the detail included in the model, and the simplifying assumptions. Lumped-mass, stick models are simpler mathematically than finite element models and usually have a smaller number of degrees-of-freedom. Lumped-mass, stick models take advantage of the judgement of the analyst as to the expected behavior of the structure. '

Modeling methods and techniques will not be d i s ~ in detail here. The ASCE Seismic Analysis Standard and Commentary [47] present many aspects of modeling and a bibliography from which additional information may be obtalned.

One point of note for shear wall structures is that over the last half of the 1980s, testing of shear walls as individual elements and as a portion of a structure assemblage has been performed. One result of these tests is the apparent reduction in stiffness from the linearly calculated values due to small cracks and other phenomenon. Reference 48 evaluated the relevant data and recommended A approaches to account for increases and decreases in stiffness previously not treated explicitly.

- Increases result from items such as increased concrete strength due to aging and achieving minimum specified strengths in a conservative manner.

234 42.2. Nonlinear Structure Models. The nonlinear behavior of structures ls important in two regards

- evaluating the capacity of snuctural members and the structure itself; and estimating the environment (In-structure response spectra and structural displacements) to which equipment and commodities are subjected. Nonlinear structural behavior is characterized by a shift in natural frequencies to lower values, increased energy dissipation, and increased relative displacement between points in the structure. Four factors determine the significance of nonlinear structural behavior to dynamic response.

Frequency content of the control motion versus the frequencies of the structull. Consider a rock-founded structure. Depending on the elastic structure frequencies and characteristics of the control motion, the shift in structure frequency due to nonlinear structural response may result in a relatively large reduction in structure response with a substantial reduction in input to equipment when compared to elastic analysis results. If the structure elastic frequency is located on the peak or close to the peak of the control motion's response spectra. the frequency shift will result in a much large decrease in response than if the elastic frequency is higher that the peak of the control motion's response spectra such that the shtft in frequency tends to result in increased response. Toe same type of reduction occurs for structures excited by narrow band spectra earthquakes where the structure elastic frequency tends to coincide closely with the peak of the ground response spectra.

Soll-structure interaction effects. The effect of nonlinear structure behavior on in-structure response (forces, accelerations, and response spectra) appears to be significantly less when SSI effects are important at the site. This is principally due to the potential dominating effect of SSI on the m,ipomtA of the soil-structure system. The soil can have a controlling effect on the frequencies of the soil.,

structure system. Also, if SSI is treated properly, the input motion to the system is filtered such that higher frequency motion is removed., i.e., frequency content which may not be suppressed by nonlinear structural behavior if the structure were founded on rock. SSI can have a significant effect on the energy dissipation characteristics of the system due to radiation damping and material damping In the soil. Accounting for the effect of the inelastic structural behavior on structure response must be done carefully for soil-founded structures to avoid double-counting of the energy dissipation effects.

Degree of structural nanlw/arlty. Toe degree of structural nonlinearity to be expected and permitted determines the adequacy assessment fur the structure and can have a significant impact on in-structure responses. Past reviews of testing conducted on shear walls has indicated that element ductilities of up to about four or five can be accommodated before significant strength degradation begins to occur. However, the allowable achieved ductility for many evaluations will be substan:tially less. The effect of nonlinear structure behavior on in-structure response spectra has been considered to only a limited extent In general, increased levels of nonlinearity lead to increasing reduced In-structure response spectra for a normalized input motion. However, in some insmnces, higher frequency, i.e., higher than the fundamental frequency, peaks can be ampllfled. This is an area for which research is currently being performed and will provide guidance in the future.

Magnitude effects. Earthquake magnitude as it affects the control motion has been discussed earlier.

Recall, however, smaller magnitude, close-in earthquakes may be narrow-banded which are A significantly affected by nonlinear behavior as described above. -

235 4.3 SSI REsPONSE Very few cases exist where the data necessary to effectively analyze a soil-structure system has been developed or measured. In addition, for those with data, typically, not all aspects of the SSI phenomenon are important One case in point is the Lotung one-quarter scale model. All appropriate data has been developed and measured but the high stiffness of the structure and the very soft soil conditions eliminate structure vioration as a significant phenomenon. F1gure 28 [9] shows a comparison of measured and calculated response which is an excellent match.

In addition to the Lotung experunent, there is a recognition in the technical community that additional data appropriate for SSI analysis methods benchmarking and development are oxessary.

An additional experiment called the Hualien Large-Scale Seismic Test for Soil-Structure Interaction Research is being constructed in Taiwan [49]. This is a stiffer site than Lotung and eliminates some of the deficiencies of the Lotung experiment Lastly, the U.S.G.S. sponsored a workshop on February 1992 wluch itemized the need for additional measurements on soil and structures for SSI analysis methods development and benchmarking.

5. Conclusions Many aspects of SSI are understood and any valid method of analysis is able to reproduce them.

~ o n s 2, 3, and 4 itemized the various aspects of the problem and one's ability to model them.

W!e!';lY, uncertainties exist in the process; randomness associated with the earthquake ground motion itself and the dynamic behavior induced in soil and structures. Even !IS.5Ullling perfect modeling, randomness in the response of structures and components is unavoidable. Perhaps the best evidence of such randomness is the Chiba Held Station. Shibata [30] reports the results of 20 years of recorded motion on the model structure; 271 events when taken in total. Analyzing the measured responses of the hanged tank (Fig. 26) yields a coefficient of variation of response of about 0.45 conditional on the horizontal peak ground accelerations of the earthquake. Of course, arguments can be made that many of the events were low amplitude, that group events by epicentral area reduced variability, etc. However, the undisputable fact is that significant variability in response of structures and components due to earthquakes is to be expected No deterministically exact solution of the SSI problem can be obtained by existing techniques.

However, given the free-field ground motion, data concerning the dynamic behavior of soil and structure, reasonable estimates can be made. In addition, uncertainties In each of the elements do not necessarily combine in such a fashion as to always ina"ease the uncertainty in the end item of interest (structure response). Reference 50 reports on a probabilistic response study for the Diablo Canyon Nuclear Power Plant which demonstrates that combining uncertainties due to the square root of the sum of squares rule is conservative and can be extremely conservative. Hence, although uncertainties exist m one's ability to represent and model the SSI phenomenon, they do not always increase the uncertainty in the end item of interest

236

6. References

1. Johnson, J.J. (1981) "Soil-Structure Interaction: The Status of Current Analysis Methods and Research, Lawrence Livermore National Laboratory (LLNL)," UCRL-53011, NUREG/CR-1780, prepared for the U.S. NRC.

2. Luco, J.E. (1980) "Linear Soil-structure Interaction," LLNL, Livermore, CA, UCRL-15272, included in Ref. 1.

3. Roesset, J.M. (1980) "A Review of Soil-Structure Interaction," LLNL, Livermore, CA, UCRL-15262, included in Ref. I.

4. Seed, H.B. and Lysmer. J. (1980) "The Seismic Soil-Structure Interaction Problem for Nuclear Facilities," LLNL, Livermore, CA, UCRL-15254, included in Ref. 1.

5. Veletsos, A.S. (1978) "Soil-Structure Interaction for Buildings During Earthquakes," in Proc.

Second International Conference on Microzonation for Safer Construction - Research and Application, San Francisco, CA, Vol. 1, pp. 111-133.

6. Chopra, A.K. (1980) Personal Communication.

7. Johnson, JJ. and Chang, C.-Y. (1991) "State of the Art Review of Seismic Input and Soll-Structure Interaction," Appendix E in "A Methodology for Assessment of Nuclear Power Plant Seismic Margin (Rev. l)," EPRI, Palo Alto, CA, EPRI NP-6041-SL

8. EPRI (1989) Proceedings: EPRI/NRC/IPC Workshop on Seismic Soil-Structure Interaction Analysis Techniques Using Data From Lotung, Taiwan, Electric Power Research Institute, Palo Alto, CA EPRI NP-6154, Vols. 1 and 2.

9. Johnson, JJ., Maslenlkov, O.R., Mraz, MJ., and Udaka, T. (1989) "Analysis ofLarge-S~--

Containment Model in Lotung, Taiwan: Forced Vibration and Earthquake RespoU.W Analysis and Comparison," included in Reference 8, pp. 13 13-44.

IO. Tseng, W.S., and Hadjian, A.H. (1991) "Guidelines for Soll-Structure Interaction Analysis,"

EPRI, Palo Alto, CA EPRI NP-7395.

11. U.S. Atomic Energy Commission (1973) "Regulatory Guide 1.60, Design Response Spectra for Seismic Design of Nuclear Power Plants", Rev. 1.

12. Newmark, N.M. and Hall, W.J. (1978) "Development of Criteria for Seismic Review of Selected Nuclear Power Plants," NUREG/CR-0098, prepared for the U.S. Nuclear Regulatory Commission.

13. Chang, C.-Y., Power, MS., Idriss, I.M, Somerville, P.G., Silva, W., and Chen, P.C. (1985)

"Engineering Characterization of Ground Motion, Task II: Observational Data on Spatial Variations of Earthquake Ground Motion," NUREG/CR-3805, Vol. 3, JXepared for the U.S.

Nuclear Regulatory Commission.

14. Kudo, K., Shima, E., and Sakane, M. (1988) "Digital Strong Motion Accelerograpb Array in Ashlgara Valley - Seismological and Engineering Prospects of Strong Motion Observations,"

Proc. 9th World Conference on Earthquake Engineering, Vol. VIII, pp. 119-124.

15. Chang, C.Y. (1992) Personal Communication.

16. Bozorgnia, Y .B. (1992) Personal Communication.

17. Gazetas, G., and Bianchini, G. (1979) "Aeld Evaluation of Body and Surface-wave Soil-amplification Theories," Proceedings of the Second U.S. National Conference on Earthquake Engineering," EPRI, Stanford University. A

18. Yanev, P.I, Moore, T.A., and Blume, J.A. (1979) "Fukushima Nuclear Power Station,W Effect and Implications of the June 12, 1978, Miyagi-ken-Oki, Japan, Earthquake," prepared by URS/John A. Blume & Associates, Engineers, San Francisco, California

237

19. Tanaka, T. et al (1973) "Observations and Analysis of Underground Earthquake Motions,"

Proceedings of the 5th World Conference on.Earthquak:e Engineering, Rome, Vol. 1, pp. 65&-

667.

20. Chang, C-Y. et al (1990) "Equivalent linear Versus Nonlinear Ground Response Analyses at Lotung Seismic Experiment Site," Proceedings Fourth U.S. National Conference on Earthquake Engineering, Palm Spd.ngs, CA

21. Chang, C-Y. et al (1991) "Development of Shear Modulus R.e.dnction Curves Based on Lotung Downbole Ground Motion Data," Proceedings Second International Conference on Recent Advances in Geotechnical Earthquake Engineering and Soil Dynamics, St Louis.

22. Cramer, C.H. (1991) "Turkey Flat, USA Site Effects Test Area, Report 6 Weak-Motion Test:

Observations and Modeling," Technical Report No. 91-1, California Department of Conservation, Division of Mioes and Geology, Earthquake Shaking Assessment Project

23. Proceedings of the International Symposium on the Effects of Surface Geology on Seismic Motions," (1992) Odawara, Japan.

24. Ml.dorlkawa, S., (1992) "A Statistical Analysis of Submitted Predictions for the Ashigara Valley Blind Prediction Test.," (Subcommittee for the Prediction Criteria of the Ashigara Valley Blind Prediction Test). Proceedings of the Inte.mational Symposium on the Effects of Surface Geology on Seismic Motion, Odawara, Japan.

25. Seale, S.H., and Archuleta, RJ. (1991) "Analysis of Site Effects at the Games Valley Downhole Array Near the San Jacinto Fault," Proceedings of the Second International conference on Recent Advances in Geotechnical Earthquake Engineering and Soll Dynamics, St Louis, Missouri, Papec No. 8.13.

26. Hadjian, A.H., et al "A Synthesis of Predictions and Correlation Studies of the Lotung Soil-Structure Interaction Experlment (1991) EPRI Report No. NP-7307-M.

27. Ishii, K., Itoh, T., and Suhara, J. (1984) "Kinematic Interaction of Soll-structure System B ~ on Observed Data," Proceedings of the 8th World Conference on Earthquake Engineering, San Francisco, Vol. m, pp. 1017-1024.

28. Brady, F.W. (1992) Personal Commuuications.

29. Shibata, H. "On the Reliability Analysis for Structural Design Including Pipings and Equipment," Presented at the Seminar on- Probabilistic Seismic Analysis of Nuclear Power Plants, January 16-19, 1978, Berlin.

30. Shibata, H. (1991) "Uncertainty in Earthquake Engineering in Relation to Critical Facmties."

Bulletin of Earthquake Resistant Structure Research Center, Institute of Industrial Science, University ofTokyo, No. 24, pp.93-104.

31. Abrahamson, NA, Schneider, J.F., and Stepp, J.C. (1990) "Spatial Varlatlon of Strong Ground Motion for Use in Soil-Structure Interaction Analyses," Proceedings, 4th U.S.

National Conference on Earthquake Engineering, Palm Springs, California.

32. Abrahamson, N.A., Schneider, J.F., and Stepp, J.C. (1991) "Empirical Spatial Coherency Functions fur Application to Soil-Structure Interaction Analysis," Earthquake Spectra, Vol.

7, No. 1.

33. Schneider, J.F., Abrahamson, N.A, and Stepp, J.C. (1992) "The Spatial Variation of Earthquakes Ground Motion and Effects of Local Site Conditions," Proceedings, 10th W odd Conference on Earthquake Engineerlng, Madrid, Spain.

34. Schneider, J.F., et al (1990) "Spatial Variation of Ground Motion from EPRI's Dense Accelerograph Array at Parkfield, California," Proceeding, 4th U.S. National Conference Earthquake Engineering, Palm Springs, California.

238

35. Hao, H., Oliveira, C.S., and Penzien, J. (1989) "Multiple-station Ground Motion Processing and Simulation Based on SMART-I Array Data," Nuclear Engineering and Design Ill, pp.

293-310.

36. Harichandran, R.S. (1988) "Local Spatial Variation of Earthquake Ground Motion,"

~ g s , Earthquake Engineering and Soil Dynamics ASCE Specialty Conference, Park City, Utah.

37. Somerville, P.G., et al (1988) "Site Specific &timation of Spatial Incoherence of Strong Ground Motion," Proceedings, Earthquake Engineering and Soil Dynamics ASCE Specialty Conference, Park City, Utah. ,

38. Loh, C.-H., and Chen, S.-J. (1990) "Direct10nality in Spatial Variation of Earthquake Ground Motion," Proceedings, 4th U.S. National Conference on Earthquake Englnee.ring, Palm Springs, California.

39. Woods, R.D. (1978) "Measurement of DynamJc Soil Properties," ASCE Specialty Conference on Earthquake Engineering and Soll Dynamics, Pasadena, California, Vol. I, pp.91-178.

4-0. Sun, Joseph I., Golesorkhi, R., Seed, H. Bolton (1988) "Dynamic Moduli and Damping Ratios for Cohesive Soils," Report No. UCB/EERC-88-15, Earthquake Engineering Research Center, University of California, Berkeley, California.

41. Seed, H.B., Wong, R.T., Idriss, I.M, and Tokimatsu, K (1984) "Moduli and Damping Factors for Dynamic Analyses of Cohesionless Soils," Report No. UCB/EERC-84/14, Earthquake Engineering Research Center, University of California, Berkeley, California.

42. Seed, H.B., and Idriss, I.M (1970) "Soil Moduli and Damping Factors for D ~

Response Analyses," Report No. EERC 70-10, Earthquake Engineering Research Cen9 University of California, Berkeley, California.

43. ASCE Committee on Reliability of Offshore Structures, Subcommittee on Foundation Materials (1979) Probability Theory and Reliability Analysis Applied to Geotechnical Engineering of Offshore Structure Foundations.

44. Johnson, JJ., Maslenikov, 0.R., and Benda, BJ. (1984) "SSI Sensitivity Studies and Model Improvements for the U.S. NRC Seismic Safety Margins Research Program," NUREG/CR-4-018, Prepared for the U.S. Nuclear Regulatory Commission.

45. illjikata, Ket al (1987) "Dynamic Soil Stiffness of Embedded Reactor Bulldlngs," S:MlRT 9 Transactions, Vol Kl, August 1987, Lausanne, Switzerland.

46. Kobayashi, T. et al (1991) "Forced Vibration Test on Large Scale Model on Soft Rock Site (Embedment Effect Test on Soil-strucnrre Interaction)," SMiRT 11 Transactions, Vol. 7, August 1991, Tokyo, Japan.

47. ASCE Seismic Analysis of Safety-Related Nuclear Structures Standards Working Group (1985) "Standard for the Seismic Analysis of Safety-Related Structures," Standard and Commentary, Draft.

48. ASCE (1992) Dynamic Analysis Committee Working Group Report on Stiffness of Concrete.

49. Tang, Y.K. et al (1991) "The Hualien Large-Scale Seismic Test for Soil-Strucnrre Interaction Research," Transactions, SMiRT 11, Vol. K, pp. 69-74.

50. Bozorki, G. et al (1990) "Review of the Diablo Canyon Probabilistic Risk Prepared for U.S. Nuclear Regulatory Commission, NUREG/CR-5726.

~enra

• 239 Table I EIBMENTS OF A SEISMIC RESPONSE ANALYSIS INCLUDING SOIL-STRUCTURE INTERACTION Spectficru:mn of the Free Field Ground Motion Cootrol point Control mooon (peak groUDd acceleratioo and response spectra)

Spatial vanauon ofmouoo (wave propagatioo roechBDisro)

Magrutode, duratioo Models of Soil and Structnrcs Soil properues Nommal properttes (low and high stram)

V ariabtlity SSI paramet=

Kinematic roteracUOn Foundatmn tmpedances Structure models Important features (torston, floor flexibility, frequency reduction with stram, etc.)

Variability (frequency and mode shapes, dampmg)

Nonltnear behavtor rta,uon SSI Analysts of Responses Table2 FREE-FlELD DOWNHOLE ARRAYS Location ImtrnmentDeoth{m) Reference Nanmasu, Tokyo -1, -5, -8, -22, -55 13 Waseda, Tokyo -1, -17, -ol, -123 13 Menlo Park, CA GL, -12, -40, -186 13 Rlcbmond Ftel.d Statton, CA GL, -145, -40 13 Ukisbima Park, Tokyo GL, -7:7, -ol, -127 13 Futtsu Cape, C1uba GL,-70, -110 13 Kannonzak1, y okosoka City GL, -80, -120 13 Tokyo lntcmaUonal Airport GL, -50; GL, -65 13 Ohgishima Stanon, Kawasaki Crty GL, -15, -38, -150 17 Earthquake Research lnstuute, GL,-82 13 Miyako, Tokyo -5, -18, -265 13 Tomak:omai, 1-lokk:ai.do GL, -30, -SU 13 Tateyama, Tokyo -26; -38, -100 13 Higashl-Matsuyama City, Sartama -1, -58, -121; 13 ji-cho, Shrznoka -36, -100; -49; -74 E

13 City, Chiba GL, -18 13 eatty,Nevada GL, -41 13 Fnk:mbuna Nuclear Power Plant GL, -50; -.5, -3, -85, -23.5 18, 19 Lotung, Taiwan, ROC GL, -6, -11, -17, -47 20, 21, 26 Turkey Flat, CA GL, -10, -20 22, 23 Asblgara Valley, Japan GL, -30, -95 23,24 Games Valley, CA GL, -6, -15, -22, -220 25

240 Figure l: Schematic Representation of the Elements of Soll-structure Interaction 0 --- - *

-*1 I'

2"

--i---+--+-l'

-~f.#--F---

0 - --

~(Hz!

Figure 2: Examples of Aggregated Ground-motion Response Spectra

241

,o- '

10

-~ 10 10 0

10' 1 P,,..1tc (1aa111-f)

Figure 3: Site-specific Response Spectrum 10--,. spectra 1a3 0G)

'-. 102 6

~

p==::

t:

CJ 10 1 s

~

...:i

- ~

r..

Q.

rt.I 100 10-11 10-2 10- 1 PERIOD (sec) 10° 10 1 Figure 4: Uniform Hazard Spectra

242 I!

• j

---~-___,_. .. l.al

,,.......,.., iU:I 102:

,.1" *...,..,,C.1' llltl

,: IQ-I V*rtt~*I ca.cw,,nt

~

AIJ ac.c 11r--u lrt11 1-, 9 *

.u l AK tr* u ** ***u.,,

0 ~t-1 lllt-- --~I f t 11 1 U:lg ~l Ul,I r..........>w:11-td Figure 5: Response s ~ of Free-field Ground Surface Mooons of the May 20, 1986 Earthqu3'9

-___....... ,.,.....,.._...~ ........

---

---~-"----

- - - t t l l c : . . l l t ~ ....... ~

_.IIM . . . . . . U l i l l l l l ~ . . . . . . .

I~.

11 J Figure 6:

• ..,.'::-----'-:"'---...........,-~~.,.1................i:1::a.--..1..-..........i..

IDustration of Effect of Earthquake Magnitude on Response Spectral Shape Obtained from Statistical Analysis

243 TP(m) KR1 100-50-0-

SCALE O 509(m)

-100-Figure 7: Ash.lgara Valley, Japan

3

,I

• h::no41..c-l Figure 8: Response Spectra at Rock and Alluvium Sites - Ashigara Valley, Japan

244

-11..nt:C...,..CJCIO~tiul f **** U:IS C--, 000 ..... kw *

• C

. ._ l

]

f

,_L__ _ _ _......__ _ _ _ _...__....:.:;"-'--::::z.---1 hnod INCi Figure 9: Response Spectra at Rock and Alluvium Sites - Ashigara Valley, Japan

- "'-'-- LITY m.DQMII Oill:tl&I ~ tlO ,_,. Ms

---

\ . . . - - - .......,_...1,1.,1a1 c..., 100 ..... 11,1,a

,.. .,.,: ....._ .... /'.

• - PvJOd.1--,)

Figure 10: Resporae Spectra at Rock and Soil Sites - Loma Prieta Earthquake

245

- Oula4i.Jn 11!1GJNit~I! t,...,,,. .a)C ..,...,.._

---

)ff'1Ml,---~611a1 C-* bdrb iI J

l 3

~.

~

N~

,- ~-* *.. \

.. - hnodl..al Figure 11: Response Spectra at Rock and Soll Sites - Loma Prieta Earthquake 600 . - - - - - - - - - - - - - - - - - ,

<{

o HORIZONTAL D X VERTICAL z

ws.oo D D

<{ 4 00 m

~ D w

>- 3 00 0

2 00

~100 I- xx XX~ 'XX )C X XX -,...X X 000 -t---.--r-.m"TTT"--,--,-r-r-r-TT,n--,-~~~rrl 0 01 0 1 1 10 PERIOD (SEC) IQ)QG:>lUliW;f'0.121 Figure 12: Ratio of Spectra at Treasure Island/¥erba Buena Island

246 500 -,-----*,-----,S-FO O OAKLAND WHARF TREASURE ISLAND ll FOSTER CllY u

W 4 00 Cy o__  ?, I\

(/) 'Pb\ I I \

__J 6 300 I!, f~

(/)

/ \

J \

LL *, l>

LL I I- 2 00

(/) II.

.'A , .

'----- *4 ,t> \ I t;:: G- l:r-6 I

'(]. 61 I

~

Q 1.00 13 .El._ /

I!)

(/)

000 +---~~~-~~~~--~~~

0 01 0 1 1 10 PERIOD (SEC) 14000 :lltJ.JJ UT011:1 Figure 13: Horizontal Spectra Amplification at Soft Soil/Stiff Soil 500-,-------,*-S_FO_ _ _ _ _ _ _ _ _ _ _ __

O OAKLAND WHARF D TREASURE ISLAND FOSTER CllY u

W4 00 ll o__

(/)

_J 6300

(/)

LL LL I- 2 00

(/)

t;::

0100

(/)

0 1 1 10 PERIOD (SEC)

Figure 14: Vertical Spectra Amplification at Soft Soil/Stiff Soll

247 1'"1-5 OHS

\

3048m I

• 10m -*

Ann!

TC&IJ'll accatal'Offlflln (a) _ _ _

OH!I OHM-DHB11- l5tl"l Om DH817-DH&<?-

(b) Cliowihda .....,...N ~

Figure 15: Location of (a) Swface and (b) Downhole Accelerographs [8]

E-W N--5 Figure 16: Comparison of Recorded and Computed Response Spectra (5%

Damping), Deconvolved with Iterated Strain-Compatible Properties, EventLSSf07 [8]

248 o~~~

~Jr~~4~=~~~

i~~ A-*--<~-"- :H :;:; ; ; ; "~]

l }~ ~ ; ; ::: "-~:::H ~.;; : , __ j

~E t 2

~~, J 2: $ \0 l't! SO E: ~ , ; ; ; . : i KIO I  ; ~ l l 5 10 20 !110 KO J~C,.,) F~('kl)

Figure 17: Comparisons of Response Spectra (5% Damping) of Recorded and Computed Ground Surface Motions Using Recorded Motioo at 17m Depth as Input Motion, Event LSST07 [20]

1~1 ;,* :t U) tn

,.rtt2

'1tt.lt.\cc:alftTI\IM

-- ,~~ ~: ~I -*- (U~TICJi

~ ~I ;l rn I

Ii*

1i

_,.._ t!

I

Ii l

1 lt u:2 Put ~IN'UlN:

I Ir Ji. ;~I ;; 1:i U2 1,,

Pnt **h*clt.y tl:i>~ICII UMMll * ......,..,_.IW *--~*IIW'"IM!--. -lk...,...

l~Mlt *-*1i.1---*l1.Mlfl..1Cw Figure 18: Peak: Value Statistics for Standard Geotechnical Model Predictions -

249

• o 1:51 115 1:51 Ell LO LO

" " ,Q KS2 EW KDZ Ell (f

• • "IN !ut:o-.cll/

F1gure 19: Response-Spectra Quartiles for Strong-Motion Standard Geotecbnlcal Model Predictions T-IIGL-1m Y

--..tl<al J

l Ju __- ' , l , 4 - - - - - . 4 l ~ - ~

-- u u F1gure 20: Observed Transfer Function Between Foundation Level Motion (-262m) of a Large-scale Below-grmmd LNG Tank and Free-field (-Im) Horizontal Motion (27)

250

( trl'LMAflQN II t====r-:::=-.Jll--1---- - *- ~'""""

p ...

Oll----+---4--l--l--l - - - lnt-1PN.-

C-,_.M

• -o, I

I

J -.. .

g

-lotMldOJ I/

I 0 ,...__.....__ _ _J....=.:i oms 01 L2 'll

,z ** t----+---l'--lc+--1---4----1

• ..__ __,___ _ _...J.:::-.

ton u 01

,__, 11 Figure 21: Comparison of Response Spectra of Accelerograms Recorded at Fmished Grade In the Free-field and at the Base of the Reactor Caisson at the Humboldt Bay Plant During the June 6, 1975, Ferndale. California Earthquake (4]

r1--

i:... ...~ "

Figure 22: Comparison of Response Spectra of Motions Recorded at the Base of the Hollywood Storage Buildlng and in the Adjacent Parldng Lot During the 1971 San Fernando Earthquake [13)

251

-- t---- ... ---+-... --1

, _ _ _ _ , . . _ ... --1

- * ---+---- ... ~ - .......

II_,

, <<)

I no ---

~ ,.., ,,,,

~ m, ffl) 240 I tOJ0,04011 ffl Figure 23: Cross Section Throogh Pleasant Valley Pumping Station (13]

oL-~~~= .....-~~~....._.~_;:::::,=--..J 10 -I

,..,,.~ __ ......

10-, 10. 10 I Figure 24: Comparison of Horizontal Response Spectra of Motions Recorded at the Switchyard and in the Basement of the Pleasant Valley Pumping Plant During the May 2, 1983 Coalinga Earthquake Main.iliock. (13]

252 1

'i r

i*

-..---*--~

- *--*IT~*-

Figure 25: Comparison of Response Spectra Adjacent Buildings With and Without A Basement. San Fernando Earthquake [13] W L

Figure 26: Cross-section Through Cllemical E.ngineering Plant Model, Chiba Field Station [29]

253 Figure 27: Isometric View of One-quarter Scale Model Showing Response Location and Station Codes i"~i~

• • *:_..;* is' ... ... ... ..

I.. *-~* I**------'----

tl'~

1~:

I i i

I (1) Sutt*f"-lS (bl luttoaf4U

---

~- ... ---

Figure 28: Recorded and Calculated In-structure Response Spectra, Base (F4LS) and Top (F4US), May 20, 1986 Earthquake [9]

254

'\*0 j

=-*~.,., ~

-~--'~:<:!-?'.~ ~ - -

10 20 ~ IQ ~ :i ]0 J:;; ~ !10 FREOU[H:C'T 1 '"XI r~Q!J[t,C'Y (t'::,

Coherenc,aa from LSST Event 12. !Al SaperatJOn Olstancaa of 6-1 0 Meten. 1B1 8eperat10n Dlatancff of 60-70 Metani.

u

,tu.a.wa. 0-10

• I ,

i i

10 Moan Coherencln from each of the 15 LSST Events. (A)

Seplflil1IOll Dllltl.lnaaa of 6-10 Meter&. IBI 8epmnion Olstancea of 60-70 Meteni.

Figure 29: Variation Coherency with Separation Distance (31]

Figure 30: Coherency Estimates and Smoothed Nonparametric CUrves for 4 Hz Frequency Bands (A) [34]

255 1.0 0.8

. ~ 0.6 E

c.,

c., 0.4 0.2 0.0 10 ...

Shear Strain, percent Figure 31: Normalized Modulus Reduction Relationship for Clays with Plasticity Index Between 40 to 80 [40]

40

.,C u

L.

30 0.

0-

J 0 20 ct

01 C

a. 10 E

0 Cl 0

10 ... 10 _, 10-* 10 _, 10 Shear Strain, percent Flgure 32: Strain Dependent Damping Ratios for Clays [40]

256 LEGEND

• - - - BE

,,_.......... OHSAKI t- + - .,,. TA..IMI

,---*--- CflBll a - a BASLER HOFFMAN

-50 ... EOE

--. .I

+---

BECHTEL IMPEI..L

---

'i ..__... SAOOelT & WNllY

--t- ,,A* :._ ___ _ --- _..,. ucso DEPTH (ft) -100  !'

-150

\.1 ~-fr*

I ~ *:!

0 500 1000 1l500 2000 2500 3000 SHEAR MOOULUS ("91)

Figure 33: Variability of Equivalent Llnear Shear Modulus Due to Different SSI Analyses, May 20, 1986, Lotung, Taiwan LEGEND

--

• BE Ot.Pll- (ft) -100 * - - - - CREPI BASI.EA HOFFWN

-

• EOE

- - BECHTEL

---

IMPEU 150

______. SARGENT & LUNDY

• - -~ ucso 10 16 20 DAMPING (parcanl)

Figure 34: Variability of Equivalent Linear Damping Due to Different SSI Analyses, May 20, 1986, Lotung, Taiwan

257 x1o8t/m x1olltm/rad 2*0 .:.=..MathodA ----:a 1.5 ------

--MathodA*-- B

---C 1 -----D - -- C; 1*****- 0

- ~-L-----+-

j

,~

O f-----~~~;;;;;.:.;~~

I

-:~

Q I - - - - - - - - - ' - - '-

-2.0 -

0 2 4 6 Hz: 0 2 4

~--

' - :...--:_*::~

-2.0 0 2 4 6 Hz: 0 2 4 (a) awaying spring (b) rotational spring Figure 35: Analytical/Experimental Foundation Impedances [45) 30,----------~

Non-embedment:A I I 0 ' '

"' "'0+-----""-----"-------..1 3/4-

-o

Half-embedment

A21

• ;.~ , I

.E 10 +-----*1-;*~*,---------J

6. Full-embedment:A3

~ I o~~:::t:~~-_j 0 10 20 30 (Hz)

Figure 36: Compari&'On of Horizontal DispLacement Resonance Curves at Foundation Bottom, Forced Vibration Test [46]

258 Ackoowled&ement The authors wish to acknowledge the contributions of Dr. C-Y Oumg and Mr. Maurice S. Power of Geomatrix Consultants, San Francisco, California, USA; Dr. Norman A. Abrahamson, San Gabriel, California, USA; and Dr. YousefBowrgnia ofEQE International, San Francisco, California, USA

DOCKET NUMBER PROPOSED RULE (51 f PR .f~1 5:2.Jt 00 Yl '-I 1,,-0,_J

/1"1

~ §)

r DOCKETUJ USNHC

• 93 JUN 29 P1 2 :17 Northern States Power Company 414 Nicollet Mall Minneapolis, Minnesota 55401-1927 Telephone (612) 330-5500 April 21, 1993 Mr . R. M. Kenneally Regulatory Publications Branch DFIPS Office of Administration U. S. Nuclear Regulatory Commission Washington D. C. 20555 MONTICELLO NUCLEAR GENERATING PLANT Docket No . 50 - 263 License No . DPR-22 PRAIRIE ISLAND NUCLEAR GENERATING PLANT Docket Nos . 50-282 License Nos . DPR-42 50-306 DPR-60 Comments on Draft Regulatory Guide DG-1016

Reference:

[l] Draft Regulatory Guide DG-1016, "Nuclear Power Plant Instrumentation for Earthquakes" dated November 1992 .

Attached are our comments on the Draft Regulatory Guide DG-1016 .

If you need additional clarification to any of the comments, please contact Eric Ballou at (612) 388-1121 ext . 4529.

~if-Anderson,~~r,~-.,,

Director Licensing and Management Issues Attachment C: J E Silberg JUL 3 0 1993 Acknowledged by card ..................................

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DRAFT REGULATORY GUIDE DG-1016 Comment 1:

Concem: Listed in reference [l] are at least nine specific seismic instrumentation characteristics for the seismic trigger, recorder, and acceleration sensors. We understand that technical justification for the selection of these characteristics would not necessarily be provided in [1], however, we were curious about the selected ranges of some characteristics.

Discussion: As an exercise to satisfy our curiosity, we contacted four vendors selling seismic related monitoring equipment or parts to the nuclear industry. We asked each vendor to provide enough data about their equipment so that we could determine ranges of characteristics which may be deemed acceptable for seismic equipment.

Results; We found the following ranges between the various vendors;

a. "Time beyond last seismic trigger* varies from 1 to 90 seconds.

b. "Dynamic Range of Accelerometer Sensor* varies from 1000:1 to 100000:1.

c. "Frequency Range of Accelerometer Sensor" varies from Oto 150 Hz.

d. "Recorder Sample Rate" varies from 200 to 500.

e. "Recorder Bandwidth" varies from Oto 350 Hz.

f. "Dynamic Range of Recorder" varies from 1000:1 to 100000:1.

g. "Actuating level of Seismic Trigger" is usually defined in terms of percent of full scale.

h. "Seismic Trigger set for Threshold Ground Acceleration" includes 0.02g.

i. "Recording Time" varies from 10 minutes to 50 minutes.

Recommendation: At this time we request consideration for making the following changes;

a. Change Sections 4.5.1 and 4.6.3 to say Dynamic Range should be 1000:1 or greater.

b. Change Section 4.4 to Shorten Recording Time to a more reasonable value {See comment 2).

Comment 2:

Concern: The draft recording time requirement of greater than 25 minutes comes into question when one considers the whole idea behind the Cumulative Absolute Velocity (CAV) Methodology which is based not only on the absolute acceleration but also on the duration of the event.

Discussioni We have found nothing in common between Monitor recording time and CAV Methodology. When one r eviews the work which led to the creation of CAV Methodology, one senses that the significant majority of the ground motion amplitude with which we would need to be concerned actually occur in less than 40 seconds after the initiation of the event. If the event could last 25 minutes, the usefulness of CAV could be questionable.

From figure 4 of EPRI NP-5930 one wil l find that all six of the acceleration time history plots selected by EPRI to demonstrate CAV from review of over 300 earthquakes show ground acceleration levels tapering to less than 0.025g's before 40 seconds into the event.

Actually two of the six events having acceptable CAV's (under CAV limit setpoint of 0.16g-sec) of 0.047g-sec and 0.083g-sec lasted only 5 seconds. The other four earthquakes had CAV's which vary from 0.318g-sec to l.239g-sec (between 2 to 8 times the current acceptable CAV limit), yet each of these earthquakes tapered off before 40 seconds. This calls into question the 25 minute operating time requirement.

If we assume there is a family of earthquakes and potential aftershocks that would have reasonably significant peaks (>0.025g) somewhere between 40 and 1500 seconds after event initiation, then our complete earthquake family set has grown to the point where it dwarfs the EPRI *40 sec." earthquake family set outlined in NP-5930 to demonstrate CAV Methodology. If indeed this is the case, we may need to rethink the usefulness of CAV Methodology on the basis that the CAV limit will most likely be exceeded for most earthquake events. If this is not the case then we should lower the draft operating time requirement with technical justification that significant earthquakes do not last 1500 seconds. Just lowering the requirement to 600 seconds will save p l ants money and minimize plant pre-shutdown data assessment time.

Recommendation; Review technical justification for the draft 25 min.

recorder operating time and shorten the period which captives the total earthquake duration.

Comment 3.

Concern: Based on recorder maintenance history it may possible to minimize the required maintenance and repair procedures and maintenance durations mentioned in section 8.2.

Discussion: We have found it reasonable to verify the operability of the seismic monitor under the fo llowing schedule .

a. We check the batteries quarterly .

b. We calibrate and perform in depth functional tests annually.

c. We check the peak-recording accelerometers every refueling outage .

To date we have not found results which would call for shorting the periodicity of the surveillances or the maintenance to weekly or monthly checks . We would expect that state-of-art digital instrumentation would not require weekly or even monthly checks as described in section 8 . 2 of the draft . Based on the history of checking the current instrumentation you may find that such frequent checks add little value to overall safety readiness and operability of the instrumentation but adds to the cost of plant operations .

Recommendations:

Collect a reasonable sample of the current maintenance periodicity for seismic monitors at various operating plants.

Revise the draft to reflect maintenance periodicity based on history of performance of monitors at operating nuclear plants .

DOCKET NUMBER PROPOSED RULE Pl 5 a C51 F fl '11 y0 ;1-J 1UELECTRIC Willlam J. Cahill, Jr.

GrOllp Vi<< Pr~sitknt Regulatory Publications Branch DFIPS Office of Administration U. S. Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

COMMENTS TO SECOND PROPOSED REVISION 2 TO REGULATORY GUIDE 1.12

• NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES.*

Gentlemen:

TU Electric is pleased with the opportunity to provide comments to the Second Proposed Revision 2 to Regulatory Guide 1.12 *Nuclear Power Plant Instrumentation for Earthquakes.*

The Draft Regulatory Guide (DRG) DG -1016 proposes a substantial upgrade in seismic instrumentation over what is required in either revision 1 of Regulatory Guide 1.12 or the 1974 and 1978 versions of ANSI Nl8.5. In general, TU Electric supports the proposed upgrades as being consistent with the industry initiatives sought by EPRI, NUMARC, and the NRC to change the current analog technology to state of the art digital technology. The DRG proposes extensive use of Digital Time History Tri-axial Accelerographs instead of the current peak accelerographs and tri-axial spectrum recorders .

Digital time history accelerographs would in the future be easier to interface with other remote digital equipment. The DRG should reference some industry standard that would allow common computer interface.

The DRG *proposes to substantially increase the dyndmic range and bandwidth of the instrumentation without providing a basis for the increase. DRG DG-1016 proposes that the instrument recorders have a bandwidth from 0.20 hz to 50 hz and a dynamic range of 1000:1, yet most of the structural damage which occurs during an earthquake is attributable to the lower frequencies for which the current required range of 0.1 hz to 30 hz is adequate to cover. The higher frequencies are typically quickly attenuated over distance.

The DRG does not provide a Regulatory Analysis, deferring instead to the Regulatory Analysis performed for a proposed revision to 10 CFR Part 100 and 10 CFR Part 50. In the draft Regulatory Analysis provided for the revisions to Parts 100 and 50 to 10 CFR, the point is made that a backfit analysis is not performed because the revisions apply only to applicants for future licenses and permits. This infers that the DRG like the proposed revision to Parts 50 and 100 will app l y only to applicants for future licensees and JUL 3 O 1993 Ackno\'Jledged by card ..........................,,,,..,,

400 N. Olive Street L.B. 81 Dallas. Texas 7S201

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TXX-93152 Page 2 C?f 2 permits. The value impact statement for the First Proposed Revision 2 to Regulatory Guide 1.12 states that a backfit is required for the revision.

As the cost of implementing and maintaining the proposed instrumentation would be high -and not easily justified. TU Electric would like some assurance that a backfit will not be requ ired. TU Electric believes the upgrades could best be accomplished on a voluntary basis and as part of the industry instrumentation analog to digital initiative, when a corresponding increase in reliability will offset the cost of the upgrades.

If you have any questions please contact Jose' D. Rodriguez at (214) 812-8674.

Sincerely, William J. Cahill, Jr.

By:~~

J. S. Marshall Generic Licensing Manager JDR/grp

Regulatory Publications Branch, DFIPS Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 RE: Comments - Draft Regulatory; DG-1015,1016,1017,1018,4003; S Review Plan 2.5.2; and Proposed Rules, 10 CFR Parts , 52, and 100: all related to Seismic and Earthquake Engineering at Nuclear Power Stations

- To the Regulatory Publications Branch, DFIPS:

The Vermont Geological Survey (VGS) is in receipt of proposed changes to the following draft regulatory guides:

DG-1015, Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motion; DG-1016, Nuclear Power Plant Instrumentation for Earthquakes; DG-1017, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions; DG-1018, Restart of a Nuclear Power Plant Shut Down By Seismic Event; and DG-4003, General Site Suitability Criteria For Nuclear Power Stations.

VGS is also in receipt of:

Standard review plan 2.5.2, Vibratory Ground Motion - proposed revision 3 and, Proposed rules, 10 CFR Parts SO, 52, and 100, Reactor Site Criteria:

Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants.

As the clearing house for geological issues and information for Vermont state government, VGS herein makes comment on the technical aspects of the above seismic and earthquake engineering proposals. By making these comments, VGS is not stating support for or rejection of any future proposal for nucl_ear power development in Vermont or the Northeastern United States. The means by which electrical energy is produced for Vermont consumers and the need for new energy sources constitute larger issues that are the purview of other branches of State government. Policy decisions made by the Governor and the Vermont Legislature will guide the development.of future energy projects. Title 10, Section 248 of the Vermont Statutes gives the Legislature approval before the State Public Service Board can issue a certificate of public good for the construction of a nuclear fission plant.

VGS believes that the regulations, regulatory guides, and standard review plans are being promulgated at the same time that new designs are being developed

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Regulatory Publications Branch, DFIPS March 18, 1993 j Page 2 for smaller standardized reactors that are to incorporate the safety lessons of the last 30 years of reactor operations. Accordingly, the next generation of nuclear plants must stand up to the closest scrutiny. Any new designs should be tested for safety against the most conservative assumptions. Economics should be a secondary factor in developing new designs, if nuclear power is to be a viable energy option for the future. When new designs meet safety requirements under the most conservative assumptions, the economics of those designs should then be analyzed to determine if power at reasonable costs can be provided to the consumer. Only at this stage should further exploration of feasibili ty be pursued while considering nuclear waste issues.

Comments are offered that focus on seismic and earthquake engineering and siting issues. The parts of 00-1017 that encompass seismic and earthquake engineering issues are reviewed. 00-1018, is not reviewed here as VGS does not possess expertise in nuclear power plant operations.

DG-1015 and Standard Review Plan 2.5.2 The regulatory guide 00-1015 and standard review plan 2.5.2 are linked as they focus on probabilistic and deterministic seismic hazard analysis and the determination of controlling earthquakes. NRC poses certain questions in regards to the development of the regulatory guides and the standard review plan.

Deterministic and Probabilistic In determining the weight given to deterministic methods versus probabilistic methods, VGS would ere on the conservative side. Any method should apply the expected worse case controlling earthquake to the risk assessment.

Power plant design would subsequently be tied to preparations for the worse case ground acceleration. VGS would support keeping the deterministic analysis as the primary tool using peak accelerations derived from historical seismicity and geological, seismological and geophysical investigations. Probabilistic assessments should be used a s a check on controlling earthquakes derived from the

--I deterministic method. If there are wide differences in the results of the two

- methods, the most conservative should be used. VGS believes the most conservative will be the deterministic method using peak acceleration.

be better understood by the public who will ultimately determine the viability of committing to a new generation of nuclear power plants.

In *keeping with the above, VGS believes that the probabilistic analysis In addition, it is likely that the deterministic method and the derived results will should be decoupled from any comparison to the existing nuclear power plants.

A new generation of production facilities should stand alone against the most conservative assumptions of the present era. To compare new designs against older designs *to ensure that the design levels at new sites will be comparable to those at many existing sites, particularly more recently licensed sites* would only serve to weaken the viability of a new generation of facilities. In our view, t he public can only be presented with proposals that hold the promise of safety under the most stringent guidelines. Comparison with older plants, s ome of which may be poorly sited in terms of seismic risk, raises uncertainty when determining the viability of newer, safer designs.

Ground Displacement and Damage Draft Regulatory Guide 00-1015 states:

"Because engineering solutions cannot always be demonstrated .for the effects of permanent ground displacement phenomena, it is prudent to avoid a site that has a potential for surface deformation.*

Regulatory Publications Branch, DFIPS March 18, 1993 Page 3 .

VGS agrees with this statement, but is concerned with the way the potential for ground displacement phenomena is interpreted in the Eastern United States. In the Eastern United States, the regulatory guide states that tectonic structures at seismogenic depths apparently bear no relationship to geologic structures exposed at the ground surface. Because of this, the guide emphasizes the need to not only conduct thorough investigations at the ground surface but also to identify structures at seismogenic depths that can lead to surface faulting or folding, subsidence, ground collapse, or fault creep.

During the 1755 Cape Ann earthquake, which is the largest event recorded in New England, there was considerable tumbling of chimneys and some brick walls and many stone field fences were destroyed (Earthquake Information Bulletin, U.S.G.S. Geological Survey, November-December 1976, Volume 8, Number 9)

• This occurred during an era when most buildings were of wood frame construction.

Damage would most likely be greater to taller 20th century buildings of concrete and masonry construction under similar intensities.

Damage to brittle structures could readily occur in an event of similar magnitude even with no evidence of permanent ground displacement. This leads to the conclusion that there may be zones where permanent ground displacement would not occur, but brittle structures would most certainly fail. Any such zones should be identified and eliminated from siting consideration .

Investigations The distances outlined in the regulatory guides for the scope of regional investigations (200 miles) and geological, seismological and geophysical investigations (25 Miles) should be treated as loose guidelines. For example, in the Northeast, significant historical earthquakes are the Cape Ann event of 1755 near Boston and the 1925 La Malbaie, Quebec event. For many locations in Vermont either one of these earthquakes would fall outside the 200 mile radius, however the seismic risk posed by these historic events should be considered in any siting analysis for any location in Vermont.

A wider study should be conducted outside the 25 mile limit for geological, seismological, and geophysical investigations. Such study would be appropriate for many locations in Vermont controlled by events in the Adirondacks, Quebec and New Hampshire. Detailed investigations of any glacially induced displacement should be conducted for all locations in Vermont and New England as well as studies of faults that might have the potential to be seismogenic sources

• It is unclear in standard review plan 2. 5. 2 why the seismotectonic province surrounding the site is assumed to occur 15 km from the site. If a look at all potential earthquakes reveals more than one source, the seismotectonic province in which the site resides should be treated in a conservative fashion. This would suggest that a source within a seismogenic province in which a proposed site resides should be assumed to be closer to the site than 15 miles.

The regulatory guide DG-1015 suggests that generic studies by the Lawrence Livermore National Laboratory and the Electric Power Research Institute be used in the Eastern United States for the probabilistic seismic hazard analysis. In the absence of these documents, which were not included with the draft regulatory package, the VGS believes that site specific studies for probabilistic seismic hazard analysis should be conducted in the east as well as the west.

In the Standard Review Plan 2. 5. 2 the place for measuring free field ground motion is discussed. When there is relatively uniform soils over bedrock it is suggested to take the free field measurement on the soil surface at the top of tpe finished grade . In the case where there in one or more thin soil layers over

Regulatory Publications Branch, DFIPS March 18, 1993 Page 4 lying bedrock or in the case of insufficient recorded ground motion data, the control point is specified on the bedrock or a hypothetical outcrop.

It seems that when there is uncertainty, the analysis should be more conservative. The control point should be chosen on the less competent material rather than the hypothetical outcrop at a l ocation at the top of the competent material.

There appears to be uncertainty in the method by which horizontally propagated shear waves, compressional waves or surface waves may produce the maximum ground motion at a site. As the techniques are not well defined, NRC staff will use discretion when reviewing any method of analysis. NRC has flagged what could be a large area of uncertainty in any analysis of the controlling earthquakes by deterministic methods. Better definition of NRC review methods should be available in Standard Review Plan 2.5.2, so as to decrease the apparent uncertainty.

NRC may conduct a site visit when reviewing a deterministic analysis of ground acceleration. The Standard Review plan should commit NRC to conducting a field visit to any proposed sites to see site specific details as well as local and regional field sites relevant to the seismic hazard analysis.

DG-1016, Nuclear Power Plant Instrumentation for Earthquake*

In Nuclear Power Plant instrumentation for earthquakes, the regulatory guidance would require the nuclear power plant to have equipment and software to process data within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after an earthquake. This seems like a long time given the processing power available in today's computers. If events are on the order of minutes and recorders are required to perform a minimum of 25 minutes of continuous recording, than a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> analysis time probably could be reduced to get an earlier picture of response spectra at a facility with subsequent decisions made as to plant shutdown.

DG-1017, Pre-Barthquake Planning and Xmmadiate NUclear Power Plant Operator Postearthguak* Actions*

In Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions, the determination of plant shutdown is to be made from uncorrected earthquake records. It was found in a study of uncorrected versus corrected records that the use of uncorrected methods is conservative. If this is indeed the case, VGS concurs with NRC's recommendations.

In appendix A of 00-1017, for plants at which no instrumental data are available, plant shutdown is determined from Modified Mercalli intensity levels or Richter scale magnitude levels. Presumably, Richter scale readings would be obtained from some off-site recording device. VGS fails to see how a Modified Mercalli intensity can be determined without a plant walkdown or a survey of the surrounding conmunity~ Both these approaches to determining Modified Mercalli intensities may take unacceptable amounts of time when a quick first response is needed.

If a plant does no~ have instrument data available, then the regulatory guide essentially has failed to do its job. The guidance should be tightened to insure that seismic data is available from plant instruments. If this involves redundancy in instrumentation, it should be required such as the redundancy required for all primary safety systems.

Regulatory Publications Branch, DFIPS March 18, 1993 Page 5 DG-4003, General Site Suitability, Criteria for Nuclear Power Plant Stations This draft regulatory guide deals with issues that are not entirely geologic in nature but apply to any siting process in which geologic and seismic information are integrated. VGS herein offers comments on some of these siting issues.

As a general comment on siting there is usually a lack of uniform data for screening and site selection. The draft regulatory guide should propose methods to deal with this lack of uniformity. Without some clear understanding of the irregularities in data sets, no fair determination of acceptable sites can be made.

Ecosystems A footnote on page 4 of DG-4003 lists the kinds of data that would be needed for determining site suitability. There are several additions that should be made to this list including: Aquatic biota and Wetlands. Macroinvertebrates, non-game fish species, and wetland ecosystems should be analyzed to determine the overall health of the aquatic community so that the impact of any proposed facility can be determined. The effects of a design basis accident should be determined for all ecological communities, not only the impacts of construction and operation of the facility.

On Page B-4, the guidelines suggest that the fraction of new available water that can be diverted for cooling is in the range of 10% to 20% of flow.

Any diversion of water could threaten aquatic biota under low flow conditions.

The NRC should rethink this suggestion because the U.S. Fish and Wildlife Service has a policy on flows for sustaining aquatic habitat for fish species in New England and possibly in other parts of the country. The policy sets seasonal minimum flows. In New England these flows are .5 csm in summer, 1 csm in the fall, and 4 csm during the spring runoff period. If water is diverted to the point where less than the prescribed* amounts remain in a river for example, a project would be violating the policy.

Atmospheric Conditions The draft regulatory guide indicates that surface faulting, potential ground motion and foundation conditions (including liquefaction, subsidence and landslide potential) and seismically induced floods supersede atmospheric extremes.and dispersion as critical in determining site suitability. This may be so for the immediate choice of land for the facility, but the larger atmospheric regime in which a plant sits should not be minimized.

Apparently, NRC believes that there will continue to be routine releases from nuclear plants. VGS believes that a primary reason for new designs is to provide significant reductions in routine releases or completely eliminate such releases. Reduced or el iminated emissions would then support a decreased concern for the atmospheric regime, at least for the routine releases.

If under assumed unfavorable atmospheric conditions, the dispersion of radioactivity released following a design basis accident is insufficient at the boundary of the exclusion.area and the outer boundary of the low population zone, the proposed site would not satisfy requirements. DG-4003, however indicates that the station could include compensating engineering features to mitigate for such shortcomings. This seems contradictory. To this reviewer, the term design basis indicates that all the engineering to mitigate for such an accident would already be included in the design. So under the design basis; if a site .does not s~tisfy requirements, it should be rejected. Further engineering will not add any margin of safety.

Regulatory Publications Branch, DFIPS March 18, 1993 Page 6 Other topographical features such as hills, mountain ranges, and lake or shorelines can effect local atmospheric conditions at a site and may cause the dispersion characteristics at the site to be less favorable. As this applies to much of Vermont's terrain, the difficulties of prediction under these conditions are so noted.

Population New designs should reduce risk factors to a degree that facilities could be sited in more densely populated areas. This would more fairly share the burden of siting on communities that get t he benefits from power generation.

Based on past experience, the NRC has found that a minimum exclusion area of .4 miles and low population zone of 3 miles usually provides assurance that engineered safety features can be designed to bring the calculated do~e from a postulated accident within regulatory guidelines. No reference is given on the nature of NRC's past experience. Is this based on modeling results or analysis of major accidents such as Three Mi l e Island and Chernoble? Without the backup documentation, the efficacy of the proposed distances cannot be judged.

Flooding In the discussion of flooding it is not clear whether the probable maximum flood includes the effects of dam failures. Taking Vermont Yankee as an example, it is conceivable that a probable maximum flood could occur based on analysis of the hydrologic record. During such a flood, upstream dams could fail adding a surcharge of water from impoundment storage. The additional surcharge should be taken into account when doing any probable maximum flood analysis.

Water Availability The availability of essential water during periods of low flow or low water levels is an important initial consideration for identifying potential sites on rivers, small shallow lakes, or along coastlines. Vermont, for example, though located in a humid climate, often displays extended periods of low flow.

Vermont's mountainous terrain, covered with shallow soils, tends to create flashy basins which produce considerable runoff following storms and spring melt with periods of low flow in the warmer months. Such awareness of the possibilities for extended periods of low flow shou ld be included in any siting analysis.

Socioeconomics The regulatory guidelines indicate that communities which possess notably distinctive cultural character, such as places of historical interest or have a specialized unusual industry or avocational activity, might be excessively costly to mitigate when siting and therefore would be dropped from consideration. Such selectivity could create a situation where economically viable communities, that get the benef its of the power generation, are dropped from consideration, while poorer communities bear the burden of the siting process. This is inherently unfair and every effort should be made to treat all communities equally in the siting process.

Land Use and Aesthetics The guidelines indicate that land use plans adopted by Federal, State, regional , or local governmental entities should be examined during siting process. This should be extended to include State land use statutes as. well as s ~atutes that cover the siting of energy facilities .

Regulatory Publications Branch, DFIPS March 18, 1993 Page 7 Proposed rules, 10 CPR Parts 50, 52, and 100, Reactor Site Criteria: including Seismic and Earthquake Engineering criteria for Nuclear Power Plants.

The comments made above for the regulatory guides and standard review plan apply to the proposed regulations as well. The following are additional comments on the federal register notice containing the proposed regulations.

Siting Policy Task Force Recommendations NRC rejects policy recommendations number 6 and 8 in the *Report of the Siting Policy Task Force,

• August 1979. VGS believes that both of these recommendations have merit and should be included in the regulations.

Number 6 recommends that sites should be selected so that there are no unfavorable characteristics requiring unique or unusual design to compensate for site inadequacies. Recommendation 6 fits well with the concept of a_ smaller safer uniform design for future nuclear plant development. With a standard design, any site factor for which the design needs modification would pose a hindrance to the viability of the concept of uniformity, therefore additional engineering should not be provided for aberrations in site conditions.

Recommendation 8 proposes that a final decision by a state agency whose approval is fundamental to the project would be a sufficient basis for NRC to terminate review. The termination of a review would then be reviewed by the Commission. VGS supports recommendation 8 as the State of Vermont participates in review of federally licensed projects through federal laws that require state input such as the 401 water quality certificate. In Addition, Vermont has~

number of laws that are concerned with the environmental impact of proposed projects and the health and safety of its citizens. Where applicable, the results of State review should be given considerable weight by NRC because Vermont State Regulators would be close to information sources for any proposed project in Vermont. Proximity to the project gives a perspective on siting that NRC needs for a final siting determination.

Thank You for the opportunity to comment on the regulations, regulatory guides, and standard review plan.

Sincerely,

~~~ f<_. ~~

Laurence R. Becker Radioactive Waste Assistant LRB/lrb cc: Bill Sherman, PSD Ed von Turkovich, Emergency Management

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DOCKETED MAR 2 4 1993 j ~~s~o FOUND .E D 1891 j

!5!5 EAS T M O "N ROE ST RE ET

• 93 JU 29 P12 :1J H ICAGO, ILLINOIS 60603 c312 > 2es>>-20 o BRYAN A. ERLER PARTNER 312-269-7132  ; I Al 5f*a*~*

~ ~ch 23, 1993 The Regulatory Publication Branc~

DFIPS, Off ice of Administration

• f Qr' U. s *. Nuclear Regulatory-* _C ommissio . .. ** C{J Washington,

• D. c * . 2055_5 . . .

RE: Comments .on Proposed Rule Changes to o CFR Part 100, Appendix B; 10 .,CFR_Part .so, Appendix  ; Draft Regulatory Guides DG-1015 ," DG-1016, DG-1017, DG-1018 and SRP 2. 5. 2, Rev. 3 *  ::-* * :**

• l

Dear Reviewer:

En~losed **are -', our comin~rit~-Jon the subject proposed rule changes.

We would appreciate ~yo11r Jcfonsideration in incorporating these comments in finaliz ing *thei regulatory positions covered in the subject documents. *

, * : ** V ';i*4*i~;

Yours very truly,  ;*

B. A. . Erler .

Assistant Manager Structural Department BAE:nibl .

Enclosure

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SARGENT & LUNDY ENGINEERS CHICAGO COMMENTS ON PROPOSED RULE CHANGES TO 10 CFR PART 100, APPENDIX B; 10 CFR PART 50, APPENDIX S; DRAFT REGULATORY GUIDES DG-1015, DG-1016, DG-1017, DG-1018 AND SRP 2.5.2, REV. 3 Appendix B to 10 CFR Part 100 - criteria for the Seismic and Geologic siting of Nuclear Power Plants on or After [Effective Date of the Final Rule]

1. Appendix B requires that both deterministic and probabilistic seismic hazard evaluation must be made to assess the adequacy of the Safe Shutdown Earthquake Ground Motion.

Comment The requirement of the deterministic evaluation should be deleted. Due to inherent uncertainties in the seismic source interpretations, it is most logical to rely on probabilistic evaluations. Within the last ten years, probabilistic seismic hazard assessment methodologies have reached a level that they can be used with confidence to assess the adequacy of Safe Shutdown Earthquake Ground Motion (LLNL and EPRI methodology).

2. Under Item V (c), Determination of Safe Shutdown Earthquake Ground Motion, the last paragraph states, "At a minimum, the horizontal Safe Shutdown Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.lg."

SARGENT 8c LUNDY ENGINEERS CHICAGO comment The intent of this statement is not clear. If the intent is to have the peak of the Horizontal Safe Shutdown Earthquake Ground Motion to be at least O.lg, then it should not be referred to the foundation level. Instead it should be referred to free field ground motion at the free ground surface or hypothetical rock outcrop as appropriate. Note that in this appendix and in all other Qocuments (10 CFR Part 50 Appendix S; SRP 3.7.1, Rev. 2 and SRP 3.7.2, Rev. 2) the Safe Shutdown Earthquake is defined by free-field ground response spectra at the free ground surface or hypothetical outcrop, as appropriate. on the other hand, if the intent is to limit the reduction in the horizontal motion due to deconvolution effect in a soil-structure-interaction analysis, then 10 CFR Part 100, Appendix Bis not the place for this detail requirement. SRP 3.7.2 is an appropriate place for specifying such a limit. The SRP 3.7.2, Rev. 2, on page 3.7.2-10 has such a limit on the reduction, i.e., "The spectral amplitude of the acceleration response spectra (horizontal component of motion) in the free field at the foundation depth shall be not less than 60% of the corresponding design response spectra &t the finished grade in the free field". Provision of such a detail requirement in 10 CFR Part 100 will create a dilemma similar to the one we are facing now, i.e., inconsistent requirements in 10 CFR Part 100 and SRPs. SARGENT 8c LUNDY ENGINEERS CHICAGO Recommendation Replace the last paragraph of Section V (c) by the following:

                   "At a minimum, the horizontal Safe Shutdown Earthquake Motion at the free-field ground surface or hypothetical rock outcrop, as appropriate, must be an appropriate response spectrum with a peak ground acceleration of at least 0.lg."

Appendix s to 10 CFR Part 50 - Earthquake Engineering Criteria

...-.:-- for Nuclear Power Plants

1. Under item [l] of IV [a) Safe Shutdown Earthquake Ground Motion, it is stated: "At a minimum, the horizontal Safe Shutdown Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least 0.lg. 11 comment Comment Number 2 given above for Appendix B to 10 CFR Part 100 applies.

Draft Regulatory Guide DG-1015 - Identification and Characterization of seismic Sources, Deterministic Source Earthquake, and Ground Motion

1. Similar to Appendix B to 10 CFR Part 100, DG-1015 requires both deterministic and probabilistic seismic hazard evaluation to assess the Safe Shutdown Earthquake Ground Motion.

SARGENT & LUNDY ENGtNEERS CHICAGO Comment Comment Number 1 given for Appendix B to 10 CFR 100 applies, i.e., the requirement of the deterministic evaluation should be deleted.

2. Lines 22-25 of page B-1 states, "The following procedure is one approach acceptable to the NRC staff to assure that the annual probability of exceeding the SSE compares favorably with that for the nuclear power plants operating as of the date of the final version of Appendix B to Part 100."

comment Note the words "compares favorably" above and the words "considered acceptably low if it is less than the median annual probability" in Appendix B to CFR Part 100. The wording of acceptance criteria for the same parameter must be the same in all documents. (There is a similar wording "compares favorably", , Lines 35-36 of page 8, Lines 11-12 of page B3).

• Recommendation In Section C - REGULATORY POSITION of DG-1015, provide a statement for acceptable annual probability of exceeding the SSE, similar to the one given in Appendix B to CFR Part 100.

Specify the numerical values of lE-4 and 3E-5 for median annual probabilities obtained by LLNL and EPRI methodologies respectively for current plants in eastern U.S. sites. SARGENT & LUNDY ENGINEERS CHICAGO

3. Section B.2.2 Western U.S. Sites Comment This Section does not contain a guidance for acceptable probability level for exceeding SSE in Western sites.

Recommendation Add the following guidance:

         "A site-specific seismic hazard evaluation using EPRI or LLNL methodology is acceptable.

The accepted median annual probability levels for these methodologies are lE-4 and 3E-5 for LLNL and EPRI respectively (i.e., same as those for Eastern sites), unless specific studies are done to justify higher acceptable probability level." Draft Regulatory Guide DG-1016 - Nuclear Power Plant Instrumentation for Earthquakes

1. Line 28 page 3.

comment Delete the adjective "state-of-the-art" modifying solid-state instrumentation. This is not necessary, and it is ambiguous. SARGENT & LUNDY ENGINEERS CHICAGO

2. On pages 3 and 4 item 1.2 lists the location and numbers of j

triaxial time-history accelerographs. Comment (a) In the case of rock site, seismic motion at the containment foundation will be similar to the motion at the free field,_ hence add a note stating that for a rock site (soil-structure interaction is negligible), a single time-history accelerograph at either location 1 (free-field) or location 2 (containment foundation) may 4t be provided. (b) Item 4 of 1.2 requires accelerographs on two independent Cat. I structure foundations. Accelerograph on one independent Cat. I structure foundation is sufficient.

3. Line 35-36 on page 5, "The instrument should be capable of a minimum of 25 minutes of continuous recording".

comment;

• The 25 minutes of continuous recording time is excessive and not needed. This requirement will unnecessarily restrict the use of several accelerographs available in the market.

For example, Kinametrics accelerograph SSA-2 meets all the other requirements, but it has a recording capacity of approximately 10 minutes. It is recommended that a minimum of 10 minutes continuous recording is a reasonable desired capacity. SARGENT 8c LUNDY ENGINEERS CHICAGO

4. Lines 23-25 on page 6, 11 5.1 The instrumentation should be I

designed and installed so that the vibratory transmissibility over the amplified region of the design spectral frequency range is essentially unity, that is, so that the mounting is rigid." Comment: Replace the sentence by, "The instrument mounting should be rigid." Draft Regulatory Guide DG-1017, Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions

1. Section 4.1 Response Spectrum Check Comments:

(a) The Section does not provide a guideline for frequencies at which the 5% damped free field ground motion response should be calculated .

• (b) The double specifying optional rules within the parentheses of Items 1 and 2 of Section 4.1 could cause confusion. Per proposed 10 CFR Part 100, appendix B, applicant selects only one OBE.

Recommendation Replace the content of Section 4.1 as follows:

           "The OBE response spectrum is exceeded if any one of the three components (two horizontal and one vertical) of the 5% damped free-field ground motion response spectra calculated at
 ' .,,                           SARGENT & LUNDY ENGINEERS CHICAGO 15 frequencies approximately evenly spaced on a logarithmic scale for frequencies between 1 to 10 Hz, is larger than:

1. The corresponding OBE response spectral acceleration or 0.2g, whichever is greater, for frequencies between 2 to 10 Hz.

2. The corresponding OBE response spectral velocity or a spectral velocity of 6 inches per second, whichever is greater, for frequencies between 1 to

• 2 Hz."

Draft Regulatory Guide DG-1017 - Restart of a Nuclear Power Plant Shut DoWn by A Seismic Eyent No comments standard Review Plan 2.5.2 - Vibratory Ground Motion Proposed Revision.a 1* Similar to Appendix B to 10 CFR Part 100, this proposed

• revision requires both deterministic and probabilistic evaluations to assess the'ssE.

comment: Comment Number 1 given above for Appendix B to 10 CFR 100 applies, i.e., the requirement of the deterministic evaluation should be deleted. D C .ET r'UMBER PRuPOSED RULE R t; 0 1 S-1=,J--JOO ( 5" 7 F K!_ L/ 7 frO?..) . [ CKLi Ee US NRC L0CKi_ iLO 1 USNHC AMERICAN NUCLEAR SOCIETY *93 JUN 28 P4 :34 NUCLEAR POWER PLAP'!J SJlj'~IJ'ARDS COMMITTEE

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Walter H. D'Atdenne, PhD, PE LaGrange Park, Dlinois 60525 USA i\Ol: Ki I I '1 ,. t .. ~ '- GE Nuclear Energy 175 Curtner Avenue, MC 487 San Jose. CA 95125 June 24, 1993 Telephone: 408/925-1214 FAX: 408/925-1289 or 1150 Mr. Samual I. Chilk CC: A David Rossin Secretary of the Commission Edward D. Fuller Mail Stop 16G15 Kenneth C. Rogers US Nuclear Regulatory Commission Washington, DC 20555

SUBJECT:

Proposed Rulemaking 10CFR Parts 50, 52 and 100, "Reactor Site Criteria" (57 Federal Register 47802 - October 20, 1992 and 55601- November 25, 1992) ATTN: Docketing and Service Branch

Dear Mr. Chilk:

The Nuclear Power Plant Standards Committee (Nuppsco) of the American Nuclear Society has reviewed the proposed rule "IOCFR Parts 50, 52, and 100, Reactor Site Criteria," (57 Federal Register 47802 - October 20, 1992 and 55601 - November 25, 1992) and offers the attached comments for consideration. The comments are presented in two parts: (I) Non-Seismic Issues and (II) Seismic Issues. Nuppsco is one of several consensus standards committees sponsored by the American Nuclear Society and is responsible for the development and maintenance of American National Standards in the area of siting of nuclear power plants. The enclosed comments reflect the technical judgment of the members ofNuppsco only and do not necessarily reflect the opinions of the American Nuclear Society. We appreciate this opportunity to comment on the proposed rulemaking. Members of NUPPSCO would be happy to meet with the NRC to discuss any of these comments. Very truly yours,

                   ~~#, ,, ~

Walter H. D'Ardenne JUL 3 0 1993 Acknowledged by card ............................._ ..

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ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100 The Nuclear Power Plant Standards Committee (NUPPSCO) is one of several consensus standards committees sponsored by the American Nuclear Society and is responsible for the development and maintenance of American National Standards in the area of siting of nuclear power plants. The following comments reflect the technical judgement of the members ofNUPPSCO and do not necessarily reflect the opinions of the American Nuclear Society.

GENERAL COMMENT

As an overall comment, the proposed rule changes are more prescriptive than the current NRC regulations. Both the NRC and NUMARC have initiatives in progress to reduce the specificity ofNRC regulations and guidance. Consistent with those initiatives, the following comments are intended to make the proposed rule changes more technically sound, while providing for less prescriptive revisions. SPECIFIC COMMENTS L NON-SEISMIC ISSUES A. Population Density (10CFR Part 100.21 (b)) The proposed rule is based on Regulatory Guide 4.7 and recommendations of NUREG-0625, both of which have been superseded by the NRC safety goals policy statement. Of particular concern is the fact that proposed 10CFR 100.21 (b) places restrictions on the population density over an area within a 30 mile radius of the plant. It is recommended that population density limits be avoided in the regulation, and be specified in regulatory guides only if necessary and only after a value-impact analysis is made. B. Removal of Radiological Analysis from Siting Criteria As stated in the Federal Register Notice, one of the NRC objectives in proposing this rulemaking is to "[s]tate the criteria for future sites that, based upon experience and importance to risk, have been shown as key to protecting public health and safety". The proposed rulemaking specifies population density and minimum exclusion area siting criteria, however, experience has shown that the key criteria for siting have been the radiological consequence evaluation factors contained in the current 10CFR 100. The current criteria have been used to safely site all licensed nuclear power plants in the United States. The application of different criteria for new plants is inconsistent with the stated objective and has the potential for significant unintended impacts on public perception regarding the acceptable safety of both existing nuclear power plants as well as advanced light water reactor designs. Adoption of the Page 1 of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100 proposed population density/exclusion area siting criteria would send an inappropriate message on the acceptability of current licensed nuclear power plant sites as well as the siting of US advanced reactor designs, both in the US and worldwide. The radiological dose consequence evaluation factors in the current 10CFR 100 are the key criteria for future sites and these criteria should be maintained in the rule. C. Answen to Questions By NRC (XLA):

1. Should the Commission grandfather existing sites with exclusion area less than 0.4 miles?

Yes. The numerical limit provides guidance at the time a site is considered; but once approved, a site should never be challenged ex post facto based on later interpretation of minor technical aspects of a rule.

2. Should exclusion distance be smaller for lower power level plants?

The important issue here is the concept of an exclusion area, and not its precise size. It is likely that future power plants will be designed so that sites of considerable size are desired, even in the case of modular designs. The exclusion area distance (EAD) could be allowed to be smaller, however, depending on the power level of the plant. Fixing a minimum value for the EAD is not recommended. The regulation should provide criteria and not set specific limits. Smaller EADs could be justified for particular sites and designs based on the following argument: There are three functions for the exclusion area distance: a) To ensure that radiation doses to the public are within 10CFR 100 limits. b) To ensure that no emergency action will be necessary beyond the EAD for lifesaving purposes in the event of a design basis accident, and c) To ensure that no activity that might pose a potential hazard to the safe operation of the plant will be in the immediate proximity of the plant. Page 2 of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100 The first two functions are related to radiation dose which decreases with the power level for reactors of the same type and fuel. The third function depends on external threats to the safe operation of the plant. Some distance away from the reactor, atmospheric dispersion reduces the dose in proportion to o-n, where D is the distance and n is an exponent in the range of 1 to 2 depending on the weather conditions. Closer to the reactor, local effects, such as the building wake and release height, result in slower decrease with distance. Based on these observations, a conservative estimate of the EAD when the power decreases from 3380 MWt to a lower power can be estimated using the inverse square law. Thus, if0.4 miles is the accepted EAD for 3380 MWt, the EAD at a lower power P can be estimated using the following equation: EAD ~ 0.4 (P/3380) o.5 (1) For a 500 MWt reactor, the above equation leads to an exclusion area distance of: EAD ~ 0.4 (500/3380) o.5 ~ 0.15 mile

                                         ~  250meters The above EAD is conservative, and will lead to an off-site dose which is less than that for the larger power reactor.

The remaining question is whether this EAD provides the desired assurance that external threats to the safe operation of the plant are sufficiently distant. This assurance depends on the site specific characteristics and on plant specific safeguards which might reduce such external threats. Based on the above observations, and assuming that an exclusion area distance of0.4 miles is acceptable for 3380 MWt, a two step approach is recommended:

1) Use Equation 1 to estimate the exclusion area distance for lower power reactors.

2) Evaluate wlnerabilities to external threats based on site specific characteristics and plant specific safeguards.

Page 3 of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100

3. The commission proposes to codify Reg. Guide 4.7.

This should not be done. The Regulatory Guide itself should be rescinded as providing improper guidance and being unnecessary for site selection or evaluation. a) Numerical values of population density should not appear in the regulation. General guidance, like that currently in Part 100, has proved to be sufficient, workable and justifiable. No Regulatory Guide should be issued that contains any population density criteria. b) Numerical values: This notice asks what the basis should be for other numerical values. There is no basis for the numerical values proposed in this Notice in the first place. No basis has ever been offered that holds up under scrutiny for any population density criteria within any radius around a plant. There is no technical basis for criteria out to distances of 20, 30 or 40 miles. c) Distances: As above, there is no justification for any particular distance for setting population density criteria.

4. Future sites that might exceed population values: Questions oflarge population centers can be raised in hearings for site acceptability and considered on the basis of communications, traffic, etc. in relation to alternative sites, but not as a rigid criterion which would have to be reconsidered for otherwise acceptable sites.

5. Periodic reporting and updating: Once a site is approved, it should remain acceptable without challenge or reconsideration. If a facility, such as a plant that processes explosive chemicals, is proposed for a site near by, the licensee or the NRC should raise any question it might choose regarding that facility. The Commission needs no other involvement with the area surrounding an approved site. Neither party has jurisdiction over land use outside the plant site, and cannot be put in a position of reconsidering a long-term licensing commitment because of what other parties, fully aware of the site's existence, should choose to do.

6. Continuing obligations: 10CFR 100 governs siting. Once a plant has been sited under a Part 50 or Part 52 application, Part 100 should have no further effect.

Page4 of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100

7. Meteorological conditions: Meteorological data are required for the environmental impact statement for any site. They can be raised in the site suitability hearing. No rigid rule or requirement would make sense in light of the history of licensing and operating nuclear power plants around the world. Regulations have been adopted for hurricanes and other major storms or natural disasters.

8. Siting Policy Task Force Report NUREG-0625: As noted above, NUREG-0625 has been superseded by the NRC safety goals policy statement and therefore should not be the basis for new or revised regulations.

IT. SEISMIC ISSUES A. Use of Lawrence Livermore National Laboratory (LLNL) NUREG/CR-5250 (January 1989) At a public meeting held by the NRC in Rockville, MD on March 9, 1993 LLNL made a presentation which showed reductions in seismic hazards at nuclear power plants of one to two orders of magnitude below the values published by LLNL in NUREG/CR-5250, "Seismic Hazard Characterization of 69 Nuclear Power Plant Sites East of the Rocky Mountains." These reductions were made possible by a re-elicitation by LLNL of expert opinion in the inputs of"recurrence" and "attenuation". As a consequence of this presentation, it is clear that NUREG/CR-5250 is no longer supported by LLNL. Further, it appears from comments made by LLNL at the meeting that LLNL has no plans to publish a revision to NUREG/CR-5250 to document the reduced seismic hazards it has calculated. This being the case, NUREG/CR-5250 should not be referenced in the proposed rulemaking and its seismic hazard curves should not be used in the 10CFR 100, Appendix B site selection process. B. Use of 0.lg SSE 10CFR 100, Appendix B.V (c) and 10CFR 50, Appendix S.IV (a)(l) state: 11 At a minimum, the horizontal Safe Shutdown Earthquake Ground Motion at the foundation level of the structures must be an appropriate response spectrum with a peak ground acceleration of at least O.lg." The intent should be to have the peak of the horizontal Safe Shutdown Earthquake Ground Motion be at least O. lg, but it should not be referred to the foundation level. Instead it should be referred to free field ground motion at the free ground surface or hypothetical rock outcrop as appropriate. Note Page S of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100 that in this appendix and in all other documents (10CFR Part 50 Appendix S; SRP 3.7. 1, Rev.2 and SRP 3.7.2, Rev. 2) the Safe Shutdown Earthquake is defined by free-field ground response spectra at the free ground surface or hypothetical rock outcrop, as appropriate. The requirements should be revised to read as follows:

           "At a minimum, the horizontal Safe Shutdown Earthquake Motion at the free-field ground surface or hypothetical rock outcrop, as appropriate, must be an appropriate response spectrum with a peak ground acceleration of at least 0. lg."

If the intent of a minimum acceleration of 0. lg at the foundation level of structures is to establish a minimum design criterion for structures, systems and components regardless of the site characteristics and soil-structure interaction characteristics, then such a requirement should be incorporated into the design requirements in 10CFR 50 and not into the rulemaking for siting criteria. C. Answers to Questions Requested by NRC (XI.B)

1. Use of both deterministic and probabilistic seismic hazard evaluations.

The requirement of the deterministic evaluation should be deleted. Due to inherent uncertainties in the seismic source interpretations, it is most logical to rely on probabilistic evaluations. Within the last ten years, probabilistic seismic hazard assessment methodologies have reached a level that they can be used with confidence to assess the adequacy of Safe Shutdown Earthquake ground motion.

2. Determination of controlling earthquakes for probabilistic analysis.

The Safety Goal Policy Statement was approved by the Commission in June 1986 and published in August 1986. On June 15, 1990 the Commission published a staff requirements memorandum (SRM) which offered the following gtiidance to the NRC Staff regarding implementation of tfie Safety Goals:

• The staff should establish a formal mechanism, including documentation, for ensuring that future regulatory initiatives are evaluated for conformity with the Safety Goal.
• A core damage probability ofless than 1 in 10,000 per year of reactor operations (1 x 1Q-4) is a useful benchmark for judging regulations on accident prevention.

Page 6 of 8

, ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100

• Safety Goal objectives should be targets for generic regulatory requirements (see SECY-89-102).

Thus, a core damage frequency of Ix !~/reactor-year is "safe enough," and the NRC Staff should not impose more stringent requirements, even on future reactors, without revising the Safety Goal Policy Statement or performing a regulatory analysis under the backfit rule. The safety goal policy statement has been considered and used in the non-seismic portions of the proposed rulemaking and there are no reasons to exclude it from consideration and use in the seismic portions of the proposed rulemaking.

3. What statistical measures (eg. mean, median, 85th percentile, etc.)

should be used to determine the controlling earthquakes? Mean values are recommended because they are used in the safety goal policy statement and are preferred in PRA work. To use the seismic hazard curves to develop site specific peak ground acceleration (PGA), they must be entered at a (generic) "reference probability," which can be derived from the safety goal policy statement as follows: Assume a fragile plant having neither seismic design nor seismic capability to withstand a seismic event of any size. The probability of core damage then becomes the probability of having the seismic event. Since the safety goal policy statement sets the threshold for core damage at 1 x 10-4, then the acceptable seismic event frequency becomes 1 x J0-4, and the corresponding PGA is obtained from the site specific seismic hazard curve. If the plant is designed and constructed to withstand a seismic event of that PGA, there will be no core damage at the acceptable probability. This is the safety goal earthquake level. The generic reference probability should be derived from the safety goal policy statement rather than the methodology in the proposed rule, which ignores the safety goals and uses median probabilities developed from the un-weighted SSEs of all existing plants regardless of design, location or vintage. The proper selection of the reference probability is of critical importance because it is the one number which all sites must use to determine their site specific PGAs from their site specific seismic hazard curves. A direct link between the reference probability and the Page7 of 8

ANS NUPPSCO COMMENTS ON PROPOSED RULE CHANGES TO 10CFR PARTS 50, 52 AND 100 safety goals is both possible and desirable in order to provide legitimacy for the reference probability. Page 8 of 8

., ATOMIC ENERCV COMMISSION Telephone: 03-6427303 Fax *  : 03-6462539

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                                                  *93 JUN 17          mo:44
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JuC J\t f R-3.2-54 Ms. P. Sobel, International Porgrams Office of Governmental and Public Affairs (WFI-3H5)

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

Dear Ms. Sobel,

As has been agreed with Dr. Litai, Head of the Licensing Division at the IAEC, while visiting at your offices in Washington, I am sending you herewith a few comments on your draft regulatory guide DG-1015. The comments have been prepared by our consultant Dr. Steinberg, a professor of statistics at the Tel-Aviv University, at present on sabbatical leave in the University of Wisconsin at Madison. He is our expert for checking the probability chapter in the Shivta PSAR. Another unclear item in your new regulations is the so-called Nseismogenic province-. We hardly understand how you practically define such a province and worse then that: how do you apply that parameter into the safety analysis. Is it another expression for the old-fashioned term "'randon1 event"', previously used in the "'seismotectonic province"'? I would highly appreciate you ellaborated, learned response on the above remarks, and better off, if feasible, any summary or report of the remarks made to you by other parties. Sincerely Yours, lr; w""_:.&k.,~ t . -Y. Weiler Head, Dept. of Site Licensing Enclosed JUL 3 o 1993 YW/AS Acknowledged by card ........................ "'"'"" P.O.B. 7061, TEL AVIV, ISRAEL, ZIP CODE: 61070 i,p,1.l ,:i,:u,c ~n ,7061.,.n TELEFAX: 03-6462974 ,t,p!I~" TELEX: STPF IL 33450 :Op~ CABLES:ATOMENERGYTELAVIV ,o,p,:m .

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• I Comments on DRAFT REGULATORY GUIDE DG-1015 and STANDARD REVIEW PLAN 2.5.2 submitted by Dr. David Steinberg Department of Statistics Tel Aviv University Draft Regulatory Guide DG-1015 of the U.S. Nuclear Regulatory Commission discusses the "identification and characterization of seismic sources, deterministic source earthquakes, and ground motion." The Guide discusses the use of both deterministic and probabilistic seismic hazard analysis.

There appear to be two fundamentally new ideas in this guide with respect to probabilistic seismic hazard analysis.

1. The evaluation of the computed probabilities by comparison with probabilities for existing nuclear facilities.

2. The use of the probabilistic analysis to define a "controlling earthquake."

I will begin by addressing the comparison procedure and then will comment on the controlling earthquake. One of the difficult questions to answer in probabilistic hazard analysis is what constitutes an acceptable probability of exceeding, say, the safe shutdown earthquake ground motion. Is 10- 6 acceptable? Or 10- 5 ? How does one establish an acceptable number? Past attempts, such as the Rasmussen Report, have emphasized equating risk from a nuclear event to risks that are commonly accepted in modern life. DG-1015 rejects this strategy in favor of an alternative: equate the hazard from a new facility with the hazards computed for existing facilities, with the new facility judged acceptable if it is below the median for existing facilities. The reason for the new approach, it seems, is that the US NRC currently recognizes two legitimate approaches to probabilistic assessment of seismic hazard (one from EPRI and one from Lawrence Livermore National Laboratories), but these two methods give answers for the eastern US that differ by an order of magnitude. (See pages B-7 and B-8 of DG-1015.) One is then faced with a dilemma - which number to believe? The NRC does not answer this question. Instead, DG-1015 suggest a creative alternative: calibrate the new facility to existing facilities. Evidently, although EPRI and LLNL give quite different answers, they do agree well on the relative hazards of different sites, (i.e. on which sites have low hazard and which have high

hazard). Thus it can be expected that the two methods will agree on the relative assessment of the new facility. I am quite troubled by the disparity between the EPRI and LLNL assessments. If these two recognized methods are in such severe disagreement, can we believe the results of either? Perhaps it is not worth doing a probabilistic assessment at all. The idea that one can overcome this problem by calibrating a new facility to existing ones is not appealing. How, after all, were the existing facilities established? In part, at least, by earlier probabilistic hazard analyses. Having now concluded that we can't get reliable probabilities from these analyses, why should we believe that the probabilities used to plan the existing facilities were reasonable? And if they are not reasonable, what value is there in using them as a reference set against which to compare the results for a new facility? DRG-1015 does not explore the crucial question of why it is that these two hazard assessment schemes produce such different results. Until this question is resolved, I think that DG-1015 casts serious doubt on the value of conducting probabilistic hazard assessments. In defense of probabilistic hazard analyses, let me note that two explanations for the different results immediately come to mind. First, it may be that the two reports differ substantially in their assessment of the relevant seismic sources. Second, it may be that they are using different methodologies to process the source information. If the former conjecture is true, the situation is less severe, since the methodology may be considered to be acceptable, with the caveat that one needs good information on the sources. If, though, the reports are using fundamentally different methodologies, then it is essential to compare the two methodologies. It is impossible to comment on this question from DG-1015 alone, since no details are given on the methods. The second new idea is the computation of a "controlling earthquake" from the probabilistic assessment. The raw output of a probabilistic assessment is a graph that shows, for each accel-eration, the annual probability of exceedance. The "controlling earthquake" provides a simple summary of the output by establishing a single event (in terms of magnitude and source-site distance) that is "typical" of the probability distribution. What is the purpose of the "controlling earthquake"? There seem to be two motivations. First, the deterministic hazard assessment is also based on the idea of a "controlling earthquake", so it may be hoped that there will be a common basis for comparing the two assessments. Second, perhaps it is believed that the probabilistic assessment will be easier to understand if it is summarized in this more tangible manner. I find both of the above arguments unconvincing. The biggest problem is that the CE from the probabilistic analysis is completely artificial. There is no reason for the computed distance and magnitude to correspond to any real source zone. For example, if a site is threatened by two source zones, one near and one far, and both make nearly equal contributions to the exceedance probability at the site (for some acceleration and frequency), the controlling earthquake will be at an intermediate distance at which there is no active fault. How can this "artificial" event be of value in comparing the probabilistic analysis to the deterministic analysis? I do not see any useful information that is gained by computing the CE from a probabilistic hazard analysis.

r On a purely technical note, I have some reservations about the method for computing the CE from the probabilistic analysis. The bins should be set up more or less on a log scale, so that bins at short distance will be much narrower than bins at long distance, reflecting the fact that the impact on hazard of near-field events is much greater than the effect of distant events. In the example in DG-1015, all events of distance up to 25 km are grouped in a single bin, even though the hazards from events at the extremes may be quite different. For near-field events, a 25 km wide bin is too large. Finally, let me make some brief remarks on Standard Review Plan 2.5.2 which discusses Vibratory Ground Motion. The major change in the revised version of this document is to A drop the concept of Operating Basis Earthquake. It appears that the Controlling Earthquake is Wreplacing the OBE in some of the site evaluations. It is not clear why the NRC has decided to drop the QBE nor what will be the practical implications of this decision.

DOCKET UMBER F01 Ov ... J RULE Tl1inois State Geological Surve f Illinois Department of Natural Resources Building 1

                                                                                         £,~[gy;~nd Natural Resources 615 East Peabody Drive                                                         US NHC Champaign, IL 61820-6964 217/333-4747 FAX 217 /244- 7004                                                *93 JUN 17 A1 C :44
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Regulatory Publication Branch DFIPS Office of Administration U.S. Nuclear Regulatory Commission Washington, DC 20555 Comments refer to the documents as follows: Draft Regulatory Guide DG-1015 - IDENTIFICATION AND CHARACTERIZATION OF SEISMIC SOURCES, DETERMINISTIC SOURCE EARTHQUAKES, AND GROUND MOTION We are in agreement with the use of the term "seismic sources". Page 6 - Use of the term "Eastern United States" here and other places in the document could be confusing to the reader. The definition provided simply indicates that this includes areas east of the Rocky Mountains. We would suggest that Midcontinent and Eastern United States" better defines the 11 regions of the U. S. that are concerned with seismic safety. Page 7 Eastern United States A "short record of the historical seismicity 11 implies that other areas of the U.S. have a longer record of historical data. We suggest that the length of the historical record is not significantly different from other areas of the U.S. The significant factor is the frequency of events. We concur with the use of both the probabilistic and the deterministic analysis procedures although some definitions may lack some specificity. Page D-9 We would add to D2.4.2 Other Quantitative Numerical Methods continuing from page D-8, archeological/geological techniques based on a combination of cultural artifacts and radio-carbon dating could be added to this list. As a general comment, we would caution that a reasonable approach to siting decisions must prevail regardless of the approach. DRAFT REGULATORY GUIDE DG-1016 - NUCLEAR POWER PLANT INSTRUMENTATION OF EARTHQUAKES We have no substantive comments on this document. JUL :; 0 1993 Acknowledged by card .........- - - - Printed on recycled paper

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Page 2 - Regulatory Publication Branch DRAFT REGULATORY GUIDE DG - 1017 - PRE-EARTHQUAKE PLANNING AND IMMEDIATE NUCLEAR POWER PLANT OPERATOR POST EARTHQUAKE ACTIONS Instrumentation and planning for pre- and post-earthquake action presumes comp liance on the part of operators. The instruments must be turned on and operating properly at all times. We would emphasize compliance with the regu lations. STANDARD REVIEW PLAN 2.5 VIBRATORY GROUND MOTION PROPOSED REVISION 3 We do not have substantive comments concerned with this document. DRAFT REGULATORY GUIDE DG-4003 - GENERAL SITE SUITABILITY CRITERIA FOR NUCLEAR POWER STATIONS The document is general and we have no substantive comments. Morris W. Lei ghton , Chief Illinois State Geological Survey Copy: Dr. James Davis

OQCl(ET NUMBER PROPOSED RULE Rs- ~ V --:~ r...,

                                                                                     / {) 0
                .                        (.§1 FR 'i1 B-01-J                  , .. ~

United States Department of the Interiorr0JJ~Fr.1l 0 GEOLOGICAL SURVEY RESTON, VA 22092 *93 JUN 14 mo :oo In Reply Refer To: V Mail Stop 905 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Chilk:

The U.S. Geological Survey is pleased to provide comments (Enclosure 1) on the proposed Appendix B to 10 CFR Part 100. These comments have been discussed informally with your staff. They are based on our ongoing research programs in geology, seismology, and engineering seismology as well as the experience gained from working cooperatively with your agency for more than 20 years to implement Appendix A to 10 CFR Part 100 and to resolve critical technical issues in the siting and design of specific nuclear power plants throughout the United States. We believe that our comments on the proposed methodology in Appendix B will satisfy two primary criteria of mutual concern. They are:

1. Adequate Margin of Safety. In the Eastern United States, the details of the geologic and tectonic framework controlling the earthquake potential and the characteristics of ground shaking are less precisely known than in the Westem United States. Therefore, we recommend that the probabilistic method be given a lower "weight" than the deterministic method until the probabilistic method has been proven to assure an adequate margin of safety.

2. Incomoration of all Available Data in Siting and Design. We believe that new research results (e.g., the data from ongoing geologic, geophysical, geodetic, seismological, engineering seismology, and engineering research; and the data acquired from worldwide postearthquake investigations, especially in seismotectonic analogs of various earthquake-prone regions of the United States; etc.) should be incorporated in every siting application.

JUL 3 0 \993 Acknowledged by card ........................-**- ..

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    \
    , '\,,

Mr. Samuel J. Chilk 2 Please call Jam.es F. Devine, Assistant Director for Engineering Geology, at 703-648-4423 if you require additional clarification on any of our comments. Sincerely yours, Dallas L. Peck Director Enclosure Copy to: Mr. Lawrence C. Shao Mr. Donnie H. Krimsley

USGS Comments on Appendix B to IO CFR 100, DG-1015, and SRP 2.5.2 The consensus of the USGS reviewers is that using the proposed hybrid method in Regulatory Guide DG-1015 is acceptable under the following provisions.

1. The Standard Review Plan (SRP) must explicitly require the NRC staff to perform an independent deterministic analysis and review to evaluate the Safe Shutdown Earth-quake Ground Motion (SSE) at the proposed site. The deterministic SSE should be evaluated as in Appendix A or following the method suggested in the last paragraph below, if that method is found to be preferable by the NRC and the USGS. Ground motions for the deterministic SSE should be determined using the 84th percentile level, as specified in SRP 2.5.2. The entire evaluation of the vibratory ground motion, including the evaluations from the deterministic method, must be documented in the Safety Evaluation Repon (SER) and be available for public scrutiny. The evaluations documented in the SER should include the SSE ground motions and the magnitude and distance determined for the Safe Shutdown Earthquake from both the deterministic and probabilistic methods. We feel that the deterministic method provides a critical check on the probabilistic results and must be specifically required by the SRP.

2. M and D derived from the probabilistic method must not be used to specify dura-tion of shaking nor to select time histories to be used in liquefaction susceptibility analysis or as direct input in the design of plant components. This should be explicitly stated in DG-1015 and other appropriate documents. It is acceptable to use M and D to select representative time histories to derive the design response spectrum, as long as this response spectrum satisfies the condition specified in item 3 below. We find that M is largely controlled by the minimum magnitude and is insensitive to the level of ground motion. Therefore, it is possible that using M would result in an underesti-mation of the duration of shaking. A minimum duration of shaking corresponding to that expected for a magnitude 6 earthquake should be used, unless a larger earthquake is appropriate.

3. Ground motions and spectra derived from M and D must exceed or equal the uni-form hazard spectrum for the appropriate probability of exceedance (see point 4 below).

4. The probability of exceedance should be based on the median of RG 1.60 plants rather than the median of all existing plants.

5. The probabilistic method must contain a provision that all new, site-specific infor-mation be incorporated into the probabilistic analysis (as in step 5 of the proposed

hybrid method). The probabilistic analysis should be reviewed using deterministic regional and site investigations.

6. It should be explicitly stated in DG-1015 that the LLNL study will be updated every 10 years (or more frequently when new information warrants) to incorporate new findings on source zones, seismicity, paleoseismology, etc.

7. The Standard Review Plan must explicitly require the NRC staff to perform the pro-babilistic analyses using both the LLNL and EPRI methodologies, unless the NRC demonstrates to the satisfaction of the USGS that the use of either the LLNL or the EPRI method produces comparable results. The evaluations (including ground motions, M and D values) derived from both methods must be contained in the Safety Evaluation Report and be available for public scrutiny.

8. The NRC needs to look further into the implications of using the probability of exceedance (Pe ) derived from the 5-10 Hz frequency band to specify the Pe for the 1-2.5 Hz band. Will using the same Pe for 1-2.5 Hz as for 5-10 Hz result in sufficient conservatism at 1-2.5 Hz? Design spectra for existing plants based on the probabilistic method (frequency-independent Pe) should be calculated and compared to the spectra for the deterministic design earthquake, based on the Boore and Joyner (1991; soil) or Atkinson and Boore (1990; rock) attenuation curves for the central and eastern United States. We are particularly concerned with the degree of conservatism in the 1-2.5 Hz frequency band. The USGS reviewers need to see this comparison before agreeing to the frequency-independent Pe.

We suggest that the NRC investigates the utility of basing the magnitude of the deterministic SSE on a uniform frequency of occurrence determined from an average of expen opinions on the frequency of occurrence for the design earthquakes of exist-ing plants. This could add stability to the deterministic approach.

Specific comments on text of Appendix B, DG-1015, and SRP 2.S.2

1. Is the site amplification incorporated in the determination of the site-specific hazard curve used in step 1, Appendix C, of DG-1015? The guide should specifically state that the site amplification must be included in the detennination of the site-specific hazard curve.

2. The second sentence in the definition 0f the DSE (p. FRN-30 and DG-1015-14) should be modified to state that the DSE magnitude should, as a minimum, be the magnitude of the largest historical earthquake associated with each source. This is stated on DG-1015-5.

3. The last sentence in the definition of "seismogenic source" (p. FRN-41, DG-1015-

13) apparently requires geologic evidence of involvement in the current (Quater-nary) tectonic regime, as well as earthquakes, to characterize a source as seismo-genic. Such evidence may not be obtainable. An alternate interpretation of the sentence is that geologic evidence of involvement in the current (Quaternary) tec-tonic regime is sufficient to characterize a seismogenic source, but surely this was not the intent. The practical meaning of the last sentence in the definition must be clarified.

4. Is "tectonic surface deformation (p. FRN-41) meant to be a sub-set of "surface deformation"? If so, it should be listed and defined separately. Is gravity one of the "earth forces " included in the proposed definition of surface deformation?

5. On page FRN-43, how does one "characterize" an assessment in this context? An improvement would be to change "must be characterized" to "must be made."

6. Various references should be changed, deleted, or added. On page DG-1015-5:

Wells and others (1989) and Wesnousky (1988) do not give empirical equations and these references should be deleted. In contrast, Wells and Coppersmith (1992, Analysis of empirical relationships among magnitude, rupture length, rupture area, and surface displacement (abs): Scism. Res. Letts., v. 63, no. 1, p. 73) give the required equations. On page DG-1015-9, the Bonilla et al. reference is wrong. It should be: Bonilla, M.G., Mark, R.K., and Lienkaemper, J.J., 1984, Statistical relations among earthquake magnitude, surface rupture length, and surface fault displacement, Bull. Scism. Soc. Am., v. 74, pp. 2389-2411. On page 2.5.2-6, the reference to Wells and Coppersmith (1 992) should be included.

I Comments on DG-1016: Seismic Instrumentation The proposed instrumentation is not sufficient to identify some of the major vibration modes of the structure, such as rocking and torsion. To identify three-dimensional behavior, we need six sensors (one per component of motion) at the foun-dation level. These sensors should be placed in at least three locations, which do not lie in a straight line. The directions of the sensors should not all be parallel and should not intersect at one poinL To identify the bending and radial modes of the containment structure, we need instrumentation at three elevations (near the bottom, middle, and top of structure). To identify soil-structure interaction, the free-field sensors should be syn-chronized with the sensors in the structure.

ATTORNEYS A T LAW

• 1615 L STREET, N.W.

WASH I NGTON , 0 .C. 2 0 03 6 - 5 610 TELEPHON E : (2 02 ) 95 5 -6 6 00 FAX : (202) 872 - 0 58 1 June 1, 1993 Mr. Samuel J. Chilk Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attn: Docketing and Service Branch Re: Proposed Rule on Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power

• Plants and Proposed Denial of Petition for Rulemaking From Free Environment, Inc. et al. (57 Fed. Reg. 47,802 (October 20, 1992))

Dear Mr. Chilk:

I The law firm of Newman & Holtzinger, P.C., on behalf of clients in its International Siting Group (ISG), hereby submits the original copy, three hard copies and one electronic copy of the ISG' s comments on the Nuclear Regulatory Commission's proposed rule, "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition for Rulemaking From Free Environment, Inc. et al.," ( 5 7 Fed. Reg. 47,802 (October 20, 1992)). ~/ The ISG has the following membership: Atomic Energy of Canada, Ltd. Electricite de France I The Federation of Electric Power Companies Hokkaido Electric Power Co. Tohoku Electric Power Co. Tokyo Electric Power Co. Chubu Electric Power Co. Hokuriku Electric Power Co.

• The Kansai Electric Power Co .

The Chugoku Electric Power Co. Shikoku Elec t ric Power Co. Kyushu Elec t ric Power Co. The Japan Atomic Power Co. Taiwan Power Company. I JIJL 3 I 1993 Acknowledged by card ........................_ .......

   ~/    Me ssrs . William O. Doub and L. Manni ng Munt zing and Ms . Janet E. B. Ecker , members of the firm ,

entered notices of appe arance as counsel for members of the ISG in this rulemaking proceeding.

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• U.S. Nuclear Regulatory Commission June 1, 1993 Page 2 The ISG was formed in response to the Commission's desire to seek the views of the international community concerning the
• proposed revisions to the siting criteria. ISG Members own and operate nuc l ear power plants in ISG Member countries. Siting of nuclear power plants in ISG Member countries is governed by national nuclear safety standards, which are consistent with the nuclear safety standards of the International Atomic Energy Agency (IAEA). These international and national siting standards were
• strongl y influenced by the Commission ' s siting standards. The Commission"s p roposed revisions to its siting regulations in 10 CFR Part 100, if adopted, would resu l t in fundamental changes to the process for selecting new nuclear power plant sites.

For the reasons set forth below and in the enclosed I comments, International Siting Group Members urge the Commission to withdraw the proposed revisions to the siting criteria and terminate the rulemaking proceeding: ( 1) Adoption of the proposed revisions to the Commission ' s demographic regulations in 10 CFR I Part 100 will do major damage to the evolving international consensus on nuclear safety standards and lead to needless inconsistency between U.S. nuclear safety standards for the siting of nuclear power plants and standards of the International Atomic Energy Agency ( IAEA) and national standards in ISG Member countries. ( 2) The existing demographic regulations in 10 CFR Part 100 have worked well. Adoption of the proposed revisions is not needed to ensure

• adequate protection o f the public health and safety nor to achieve site isolation through "decoupling" of nuclear power plant siting and design. Adoption of t he proposed revisions will not provide a substantial increa se in p rotection or contribute to increased defense-
• in-depth. The adverse impacts of the proposed revisions greatly exceed their benefits.

(3) The technical basis for the proposed revisions to the Siting Criteria is inadequate, internally inconsistent and confusing .

• (4) The proposed revisions to the Seismic Criteria should not be adopted. Revi sion should await resolution of the controversy on the use of deterministic versus probabilistic methods in site selection. Any revision adopted should

NEWMAN & HoLTZINGER, P. C. U.S. Nuclear Regulatory Commission

• June 1, 1993 Page 3 meet the Commission's rulemaking objective of regulatory stability .
• (5) The Environmental Assessment prepared in conjunction with the proposed revisions is inadequate as a matter of law to support a finding of no significant environmental impact .
• (6) Finalization of the proposed revisions to the siting regulations would not be in accord with sound agency decisionmaking.

Each of the above reasons is sufficient grounds for the

• Commission to terminate the rulemaking proceeding. When taken together, the arguments are overwhelming that to proceed at this time with the siting rulemaking is contrary to sound public policy concerning the protection of the public health and safety and the environment from radiological hazard and disruptive of internationally accepted safety norms regarding the siting and I design of nuclear power plants.

William O. Doub L. Manning Muntzing J.E.B. Ecker

• cc: Chairman Ivan Selin Commissioner Kenneth C. Rogers Commissioner James R. Curtiss Commissioner Forrest J. Remick Commissioner E. Gai l de Pl anque Mr. William C. Parler, General Counsel I Mr. James M. Taylor, Executive Director for Operations Dr. Eric s. Beckjord, Director Off i ce of Nuc l ear Reactor Regulation Dr. Thomas E. Mur l ey, Director Office of Nuclear Reactor Regulation
• Mr. Carlton Stoiber, Director Office of International Programs Dr. Paul G. Shewmon, Chairman, Advisory Committee on Reactor Safeguards

I I

• INTERNATIONAL SITING GROUP (ISG)

COMMENTS ON PROPOSED REVISIONS TO U.S. NUCLEAR POWER PLANT SITING REGULATIONS June 1, 1993 Newman & Holtzing11r, P.C. 1615 L Str1111t, N. W. Suh11 1000 W11shington, D.C. 20036 (202) 955-6600

TABLE OF CONTENTS

• PAGE EXECUTIVE

SUMMARY

. . . . . . . . . . . .         ~ . . . . . . . . . . . . . . . . . . . . . . . iv I. INTRODUCTION * * . . . . * . * . . . . * * * * . . . . . . . . . * . . * . . . . * .        1 A.       Background . . . . * . . . . * * . * . . . . . . * * . . . . . . * . . . * * . . 1

1. Summary of Proposed Revisions to the Nuclear

• Regulatory Commission's Reactor Siting Regulations 1
• 2.

3. International Siting Group {ISG) Membership . . . * * * . . ISG Members Have an Interest in Participating in This Rulemaking to Revise NRC' s Nuclear Power Plant Site 3 Safety Regulations in Part 100. . . * * * . . . . . . . . . . .

• 4

4. The Commission Has an Obligation to Consider the Implications of the Proposed Revisions to Its Siting

• Regulations on International Nuclear Power Plant Safety and Siting . . . . . * . . . . . . . . . . . . * * . . . . . * . 6

8. Summary of Reasons for Requesting Termination of the Rulemaking Proceeding to Change the Commission's Siting Regulations in 10 CFR Part 100 * * * . . . . . * * . . . . . . . . . . 10

11. MAJOR ARGUMENTS AGAINST PROPOSED CHANGES TO U.S.

NUCLEAR POWER PLANT SITING REQUIREMENTS . . . . . . . . . * . 12

• A. Adoption of the Proposed Revisions to the Commission's Demographic Regulations in 10 CFR Part 100 Will Do Major Damage to the Evolving International Consensus on Nuclear Safety Standards and Lead to Needless Inconsistency Between U.S. Nuclear Safety Standards for the Siting of Nuclear Power Plants and Standards of the International
• Atomic Energy Agency (IAEA) and National Standards in ISG Member Countries. The Existing Demographic Regulations in 10 CFR Part 100 Have Worked Well. Adoption of the Proposed Revisions Is Not Needed to Ensure Adequate Protection of the Public Health and Safety Nor to Achieve
• Site Isolation Through "Decoupling" of Nuclear Power Plant Newman & Holtzlnger, P.C.

Siting and Design. Adoption of the Proposed Revisions Will Not Provide a Substantial Increase in Protection or Contribute to Increased Defense-In-Depth. The Adverse Impacts of the Proposed Revisions Greatly Exceed Their Benefits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12

1. Adoption of the proposed revisions to the Commission's demographic regulations In 10 CFR Part 100 will do major damage to the evolving International consensus on nuclear safety standards and lead to needless inconsistency between U.S. nuclear safety standards for the siting of nuclear power plants and standards of the International Atomic Energy Agency (IAEA) and national standards in ISG Member

• 2.

countries . . . . . . . . . . . . . * . . . . . . . . . . . . . . . . . The existing demographic regulations in 10 CFR Part 100 have worked well. Adoption of the proposed revisions is not needed to ensure adequate protection 13 of the public health and safety nor to achieve site isolation through "decoupling" of nuclear power plant siting and design. Adoption of the proposed revisions will not provide a substantial increase in protection or contribute to increased defense-in-depth. The adverse impacts of the proposed revisions greatly exceed their benefits. . . . . . . . . . . . . . . . . . . . . . . . . . . *. . . . . . 21

a. The existing demographic regulations in 10 CFR Part 100 have worked well . . . . . . . . . . . . . . 21

b. Adoption of the proposed. revisions is not needed to ensure adequate protection of the public health and safety . . . . . . . . . . . . . . . . . 31

• c. Adoption of the proposed revisions is not needed to achieve site isolation through "decoupling" of nuclear power plant siting and design . . . . . . . * . . . * . * * . . . . . . . . . . . . . . 37 I d. Adoption of the proposed revisions to the demographic regulations in Part 100 will not provide a substantial increase in protection nor contribute to increased defense-in-depth. The adverse impacts of the proposed revisions
• greatly exceed their benefits . . . . . . . . . . . . . . 42 Newman & Holtzlnger, P.C.

I

B. The Technical Basis for the Proposed Revisions to the Site Safety Criteria Is Inadequate, Internally Inconsistent and Confusing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 I C. The Proposed Revisions to the Seismic Criteria Should Not Be Adopted. Revision Should Await Resolution of the Controversy on the Use of Deterministic Versus Probabilisti.c Methods In Site Selection. Any Revision Adopted Should Meet the Commission's Rulemaking Objective of Regulatory Stability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 61

1. The Commission should resolve the controversy between use of deterministic versus probabilistic

• techniques before proceeding to rulemaking . . . . . . . 62

2. The proposed requirements are unlikely to meet the stated objectives of greater predictability and stability. . . . . . . . ., . . . . . . . . . . . . . . 64

• 3. Any revision to the Commission's seismic regulations should provide regulatory stability . . . . . . . . . . . . . . 66 D. The Environmental Assessment Prepared in Conjunction with
• the Proposed Revisions Is Inadequate as a Matter of Law to Support a Finding of No Significant Environmental Impact 69 E. Finalization of the Proposed Revisions to the Siting Regulations Would Not Be in Accord With Sound Agency De_cisionmaking . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 78

 , Ill. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .            81 APPENDIX . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * . . . . . . . . . . . . . A 1 I

I I Newman & Holtzinger, P.C. Page/ii

• 'EXECUTIVE

SUMMARY

The law firm of Newman & Holtzlnger, P.C., on behalf of clients in its International Siting Group (ISG), hereby submits comments on the Nuclear Regulatory Commission's proposed rule, "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial I of Petition for Rulemaking From Free Environment, Inc. et al.," (57 Fed. Reg .

• 47,802 (October 20, 1992)). The ISG has the following membership:

Atomic Energy of Canada, Ltd . Electricite de France The Federation of Electric Power Companies I Hokkaido Electric Power Co. Tohoku Electric Power Co. Tokyo Electric Power Co. Chubu Electric Power Co. Hokuriku Electric Power Co.

• The Kansai Electric Power Co .

The Chugoku Electric Power Co. Shikoku Electric Power Co .

• Kyushu Electric Power Co.

The Japan Atomic Power Co. Taiwan Power Company . Patp/v

ISG Members own and operate nuclear power plants in ISG Member countries. Siting of nuclear power plants in ISG Member countries is governed by national nuclear safety standards, which are consistent with the nuclear safety standards of the International Atomic Energy Agency (IAEA). These international and national siting standards were strongly influenced by the Commission's siting standards. The Commission's proposed revisions to its site safety regulations in 10 CFR Part 100, if _adopted, would result in fundamental changes to the process

• for selecting new nuclear power plant sites. The fundamental nature of the changes likely would force reconsideration of IAEA and national nuclear safety siting standards and raise questions in ISG Members' countries concerning the adequacy of present and future nuclear power plant sites to ensure adequate protection of the public health and safety. ISG Members ask that the proposed
• revisions be withdrawn.

The proposed revisions are not necessary to achieve siting of nuclear power plants in areas of lower population density and away from population centers and they are inconsistent with the internationally accepted principle of establishing site safety standards which permit (and recognize the necessity to

• have) flexibility in balancing the various factors Important to the safe siting of nuclear power plants. If adopted, the regulation could force review of the presently accepted site safety principles and raise questions about whether
• presently operating nuclear power plants provide adequate protection of the public and environment when the plants are located in more densely populated areas or
• have smaller exclusion areas than the revised criteria would permit. Moreover, Newman & Holtzlnger, P. C. Pagev

should these propos~d revisions become the norm, they could preclude the siting of nuclear power plants in many areas of Western Europe and Asia and result in I a dependence on energy alternatives with less favorable environmental impacts. ISG Members believe that the NRC should consider the international

• implications of the proposed revisions to the Commission's siting regulations. A fundamental result of international nuclear cooperation has been an increased appreciation for safety standards that are shared by the entire international
• community. For the U.S. to develop and promulgate new site safety regulations.

without an appreciation for the international nuclear standards could imply a

• repudiation of current international safety standard development efforts. The proposed numerical criteria would be grossly limiting, unnecessarily so because the reviews required by the NRC and regulatory bodies in other countries . using
• standards which are reflective of international siting norms result in adequate protection of the public health and safety and in the selection of sites which are among the bes~ reasonably to be found after balancing the site characteristics important to adequate protection of the public and the environment. Likewise, decoupling of siting criteria from source terms and dose calculations to achieve site
• isolation would be entirely unsatisfactory as it would eliminate. a key measure of merit of the site-plant combination, would prevent the advantageous utilization of special design provisions in siting decisions and could provide a disincentive to
• improvement of plant safety design features. ' Such a change would be extremely limiting and certainly the wrong approach in countries or regions of countries where siting options are limited. As to the seismic criteria being proposed for Newman & Holtzlnger, P.C. P-,,evl

codification, adoption seems premature, making them unsuitable to serve as the basis for an international safety standard. Moreover, the division within the NRC

• Staff and among its experts concerning the use of probabilistic versus deterministic evaluation techniques illustrates well that the criteria do not embody the
• consensus associated with international safety standards.

For the reasons set forth in the main text of the ISG comments, International Siting Group Members urge the Commission to withdraw the

• proposed revisions to the siting criteria and terminate the rulemaking proceeding.

The principal reasons for this position include the following: I ( 1) Adoption of the proposed revisions to the Commission's demographic regulations in 10 CFR Part 100 will do major damage to the evolving international consensus on nuclear safety standards and lead to needless inconsistency between U.S. nuclear safety standards for the siting of nuclear power plants and standards of the International Atomic Energy Agency (IAEA) and national standards in ISG Member countries. (2) The existing demographic regulations in 10 CFR Part 10_0 have worked well. Adoption of the proposed revisions is not needed to ensure adequate protection of the public health and safety nor to achieve site isolation through "decoupling" of nuclear power plant siting and design. Adoption of the pr9posed revisions will not provide a substantial increase in protection or contribute

• to increased defense-in-depth. The adverse impacts of the proposed revisions greatly exceed their benefits.

(3) The technical basis for the proposed revisions to the Site Safety Criteria in 10 CFR Part 100 is inadequate,

• internally inconsistent and confusing.

(4) The proposed revisions to the Seismic Criteria should not be adopted. Revision should await resolution of the controversy on the use of deterministic versus probabilistic methods in site selection. Any revision Newt'l'IMJ & Ho/tzfnlltll', P.C. Pt,gt,vl/

adopted should meet the Commission's rulemaking objective of regulatory stability. (5) The Environmental Assessment prepared in conjunction with the proposed revisions is inadequate as a matter of law to support a finding of no significant environmental impact .

• (6) Finalization of the proposed rev1s1ons to the siting regulations would not be in accord with sound agency decisionmaking.

Each of the above reasons alone is sufficient grounds for the

• Commission to terminate the rulemaking proceeding. When taken together, the arguments are overwhelming that to proceed at this time with the siting I rulemaking is contrary to sound public policy concerning the protection of the public health and safety and the environment from radiological hazard and disruptive of Internationally accepted safety norms regarding the siting and design
• of nuclear power plants .

Newman & Holtzinger, P.C.

I. INTRODUCTION

• A. Background

1. Summary of Proposed Revisions to the Nuclear Regulatory Commission's Reactor Siting Regulatlons On October 20, 1992, the Nuclear Regulatory Commission

• (Commission or NRC) published-in the Federal Regjster (57 Fed. Reg. at 47,802) a proposed rule to change the reactor site safety requirements in 10 CFR Part 100 I (Part 100) to include specific numerical demographic requirements and to revise the seismic and geologic siting criteria in use since 1972.

The proposed changes to the site safety regulations in Part 100 I concerning demographics would set a minimum distance for an exclusion area surrounding a nuclear power reactor at 0.4 miles (640 meters). The requirement I for a low population zone surrounding the exclusion area would be deleted from the present Part 100 on the basis that the required Emergency Planning Zone (EPZ) and the proposed population density requirements obviate the need for a low population zone requirement. The proposed regulation would codify in Part 100 (the site safety regulation) the population density limits currently provided as guidance in Regulatory Guide 4. 7 in connection with consideration of alternative I sites. There would be a critical loss of necessary flexibility in making site safety determinations. Maximum population density at the time of initial site approval

• would be 500 people per square mile averaged out to 30 miles. The projected population density 40 years after initial site approval could be no more than 1000 people per square mile averaged out to 30 miles .

Newnum & Holtzlngttr, P.C.

I In addition to these numerical changes in the site safety regulations concerning demographics, a number of other revisions are proposed. Some of

• these proposed revisions include the deletion of meteorological factors from radiological dose calculations for siting purposes; modification of hydrological
• factor requirements; the addition of review of nearby industrial and transportation facilities; and the addition of periodic reporting requirements for non-related activities.

The proposed revisions to Part 100 concerning seismic and geologic siting criteria for nuclear power plants are intended to reflect advances in the earth

• sciences and in earthquake engineering. Under the seismic portion of the regulation, Safe Shutdown Earthquake (SSE) Ground Motion and site suitability criteria would be separated from design-related criteria, and detailed seismic
• guidance would be removed from the regulation. The regulation would require both probabilistic and deterministic evaluations to determine site suitability, including an explicit criterion that the probability of exceeding the SSE at a proposed site must be lower than the median annual probability of exceeding the SSE for the current generation of operating plants. In addition, the SSE calculation
• assumptions would be revised to decouple the Operating Basis Earthquake (OBE}

from the SSE. Finally, the proposed regulation would require plant shutdown in the event of vibratory ground motion in excess of the QBE. I NttWm1111 & Holtzinpr, P.C.

2. International Siting Group (ISG) Membership The law firm of Newman & Holtzinger, P.C., on behalf of clients in its

• International Siting Group USG), hereby submits comments on the Nuclear Regulatory Commission's proposed rule, *Reactor Site Criteria; Including Seismic I and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition for Rulemaking From Free Environment, Inc. et al.," (57 Fed. Reg. at 47,802 (October 20, 1992)).1' The ISG has the following membership:

I

 **                   Atomic Energy of Canada, Ltd .

Electricite de France The Federation of Electric Power Companies Hokkaido Electric Power Co. Tohoku Electric Power Co .

• Tokyo Electric Power Co.

Chubu Electric Power Co. Hokuriku Electric Power Co. The Kansai Electric Power Co. The Chugoku Electric Power Co .

• Shikoku Electric Power Co.

Kyushu Electric Power Co. The Japan Atomic Power Co .

• Taiwan Power Company.
• Menl'8. William 0. Doub end L Menning Muntzing and Ma. Janet E.B. Eoker, membeni of the firm, entered notloes of appearanoe ea counsel for m&mbere of the ISG In thia rulemaklng proceeding.

Newman & Holtzinp,r, P.C. I

The ISG was formed In response to the Commission's desire to seek the views of the international community concerning the proposed revisions to the

• site safety criteria. .SU Staff Requirements Memorandum, "SECY 92-215 --

Revision of 10 CFR Part 100, Revisions to 10 CFR Part 50, New Appendix B to 10

• CFR Part 100 and New Appendix S to 10 CFR Part 50" (August 18, 1992). In order to ensure that the comments of ISG Members on the proposed changes would be fully considered by the Commission in considering the disposition of the

' proposed changes, Newman & Holtzinger filed a request for an extension of the public comment period to June 1, 1993. On March 22, 1993, the Commission

• approved the extension request .

3. ISG Members Have an Interest in Participating in This Rulemaking to Revise NRC's Nuclear Power Plant Site Safety Regulations in Part 100.

' ISG Members own and operate nuclear power plants in ISG Member countries. Siting of nuclear power plants in ISG Member countries is governed by national nuclear safety standards, which are consistent with the nuclear safety standards of the* International Atomic Energy Agency (IAEA}. These international and national siting standards were strongly influenced by the Commission's siting

• standards. The Commission's proposed revisions to its site safety regulations in 10 CFR Part 100, if adopted, would result in fundamental changes to the process for selecting new nuclear power plant sites. Specifically, adoption of the proposed
• revisions to the site safety requirements in Part 100 concerning demographics would change demographics from one of several balancing factors to be Newm11n & Holtzinger, P.C.

considered in selecting a nuclear power plant site to a screening factor to be applied first before taking into account the other safety-related site characteristics I to be evaluated during the site selection process. These other safety-related site characteristics would be relegated to secondary importance in the site selection I process, as they would be considered only in connection with sites which first met the demographic requirements. The fundamental nature of the changes likely would force reconsideration of IAEA and national nuclear safety siting standards and raise questions in ISG Members' countries concerning the adequacy of present and future nuclear power plant sites to ensure adequate protection of the public

• health and safety .

As discussed below, the proposed revisions are not necessary to achieve siting of nuclear power plants in areas of lower population density and

• away from population centers and they are inconsistent with the internationally accepted principle of establishing site safety standards which permit (and recognize the necessity to have) flexibility in balancing the various factors important to the safe siting of nuclear power plants. If adopted, the regulation could unnecessarily force review of the presently accepted site safety principles and raise questions about whether presently operating nuclear power plants provide adequate protection of the public and environment when the plants are located in more densely populated areas or have smaller exclusion areas than the
• revised criteria would permit. Moreover, should these proposed revisions become the norm, they could preclude the siting of nuclear power plants in many areas of NIIWtlWI & Holtzlnger, P.C.

Western Europe and Asia and result in a dependence on energy '!lternatives with less favorable environmental impacts.

4. The Commission Has an Obligation to Consider the Implications of the Proposed Revisions to Its Siting Regulations on International Nuclear Power Plant Safety and Siting .

• The International Atomic Energy Agency Participation Act of 1957 (IAEA Act), 71 Stat. 453, provides in Section 3 that, "The participation of the United States in the International Atomic Energy Agency shall be consistent with I

an'd in furtherance of the purposes of the Agency set forth in its statute and the policy concerning the development, use and control of atomic energy set forth in

• the Atomic Energy Act of 1954 as amended." In conducting the instant rulemaking, the NRC thus far has failed to take into adequate account the relationship of its actions with those of the IAEA. ISG Members believe that I

pursuant to the IAEA Act, the NRC should consider the implications of the proposed revisions to the Commission's site safety regulations on the IAEA's guidelines and process with respect to siting. In order to assist the Commission in this consideration, the ISG is submitting comments regarding such implications. The IAEA, an autonomous member of the United Nations family of I organizations, came into being when its statute entered into force on July 29, 1957. As a member of the IAEA, the United States subscribes to the IAEA's objectives as defined in its statute. One of these objectives is "[to] encourage and I assist research on, and development and practical application of, atomic energy for peaceful uses throughout the world." Among the IAEA's activities related to

• safety in atomic energy are development of standards, regulations, codes of Newman & Haltzlnger, P.C. Pltge6

practice and recommendations concerning specific radiological rules, emergency procedures, and other matters. The IAEA has long considered siting of nuclear installations to be an important matter. The United States participated in a symposium held in Vienna, Austria, in December 1974 under the joint sponsorship of the IAEA and the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD) ..ll Several papers were submitted on behalf of the United States concerning its practice with respect to such matters as acceptability of nuclear sites and environmental protection. Other nations also presented papers

• and participated in discussions, leading to a rather comprehensive discussion of the matter which was published by the IAEA in 1975. Over the years, the IAEA has published a series of standards concerning the siting of nuclear facilities. For
• example, the International Nuclear Safety Advisory Group (INSAG) included siting safety principles in its Safety Series No. 75-INSAG-3, "Basic Safety Principles for Nuclear Power Plants" (1988). A "Code on the Safety of Nuclear Power Plants:

Siting" was prepared in 1978 and revised in 1988 (Siting Code). Thirteen titles in the Nuclear Safety Standards (NUSS) series concern siting .

• The extension of the public comment period to June 1, 1993, permits the NRC to satisfy its international obligations under the IAEA Act by taking into account the implications for international siting and safety standards and the role
• of the United States in their development. Although not every NRC rulemaking Sjtjng of Nyc(ear Facilities. Proceedings of the symposium held in Vienna. Austria,
• December 9-13. 1974, Jointly Organized by the IAEA and NEA (OECD).

Newman & Holtzinger, P. C*

need be accompanied by a detailed review of foreign policy implications, there are

• certain NRC rulemakings having such an obvious and direct bearing on international I

nuclear matters that failure to consider United States rules in the context of a global nuclear regime is contrary to the basic spirit of United States participation

• in the IAEA. The conduct of the instant rulemaking proceeding is one such activity.

United States interest in the .IAEA received renewed attention in the

• Nuclear Non-Proliferation Act of 1978 (NNPA). As stated in Section 201 of the NNPA, the United States is "committed to a strengthened and more effective
• International Atomic Energy Agency and to a comprehensive safeguards system administered by the Agency to deter proliferation." In addition to safeguards matters, the NNPA recognizes a number of other important roles for the IAEA. In
• particular, Section 104 of the NNPA specifies a number of desired international undertakings to be accomplished with other nations and "groups of nations such as the IAEA." _Given the primacy of the IAEA in today's nuclear society, the United States should act with a careful regard to IAEA activities. Otherwise, the United States may be perceived as undermining the effectiveness of the IAEA .
• The present rulemaking has not yet dealt with the subject of IAEA activities in the siting of nuclear facilities. In considering whether to adopt the proposed revisions, the NRC should give appropriate consideration to the role of the IAEA .
• The NRC has been directed by Congress in the NNPA and other acts to give suitable consideration to the vitally important matter of nuclear non-
• proliferation. In this rulemaking, the NRC should consider the two subjects of

siting of nuclear facilities and International impacts of NRC rules, which are obviously related to nuclear non-proliferation .

• United States foreign policy recognizes the place of nuclear energy in the economies of both the developed and the developing countries of the world.

I The safety standards and regulations developed by the NRC for the U.S. civilian nuclear industry continue to serve as models for the development of international nuclear safety standards. The proposed revisions to the site safety criteria I

 - threaten to have an adverse effect on international nuclear cooperation and to disrupt the evolving international consensus on nuclear safety standards.

• A fundamental result of international nuclear cooperation has been an increased appreciation for safety standards that are shared by the entire international community. For the U.S. to develop and promulgate new site safety
• regulations without an appreciation for the international nuclear standards could imply a repudiation of current international safety standard development efforts.

The proposed _numerical criteria would be grossly limiting, unnecessarily so because the reviews required by the NRC and.regulatory bodies in other countries using standards which are reflective of international siting_norms result in adequate

• protection of the public health and safety and in the selection of sites which are among the best reasonably to be found after balancing the site characteristics important to adequate protection of the public and the environment. Likewise,
• decoupling of siting criteria from source terms and dose calculations to achieve site isolation would be entirely unsatisfactory as it would eliminate a key measure of
• merit of the site-plant combination, would prevent the advantageous utilization of Newman & Holtz/ngfJI', P.C. Page 9

special design provisions in siting decisions and could provide a disincentive to improvement of plant safety design features. Such a change would be extremely

• limiting and certainly the wrong approach in countries or regions of countries where siting options are limited .
• As to the seismic criteria being proposed for codification, adoption seams premature, making them unsuitable to serve as the basis for an international safety standard. Moreover, the division within the NRC Staff and among its experts concerning the use of probabilistic versus deterministic evaluation techniques illustrates well that the criteria do not embody the consensus associated with International safety standards.

B. Summary of Reasons for Requesting Termination of the Rulemaking Proceeding to Change the Commission's Siting Regulations in 1Q CFR Part 100

• Based on the foregoing discussion and for the reasons set forth below in Section II, International Siting Group Members urge the Commission to withdraw the proposed revisions to the siting criteria and terminate the rulemaking proceeding. The principal reasons for this position include the following:

{ 1) Adoption of the proposed revisions to the Commission's demographic regulations in 10 CFR Part 100 will do

• major damage to the evolving international consensus on nuclear safety standards and lead to needless inconsistency between U.S. nuclear safety standards for the siting of nuclear power plants and standards of the International Atomic Energy Agency (IAEA) and national
• standards in ISG Member countries.

(2) The existing demographic regulations in 10 CFR Part 100 have worked well. Adoption of the proposed revisions is not needed to ensure adequate protection of the public health and safety nor to achieve site isolation N*wman & Holt:zlnger, P.C. Pttge10

through "decoupling" of nuclear power plant siting and design. Adoption of the proposed revisions will not provide a substantial increase In protection or contribute to increased defense-in-depth. The adverse impacts of the proposed revisions greatly exceed their benefits. (3) The technical basis for the proposed revisions to the Site Safety Criteria in 10 CFR Part 100 is inadequate, I internally inconsistent and confusing. (4) The proposed revisions to the Seismic Criteria should not be adopted. Revision should await resolution of the controversy on the use of deterministic versus I probabilistic methods in site selection. Any revision adopted should meet the Commission's rulemaking objective of regulatory stability. (5) The Environmental Assessment prepared in conjunction I with the proposed revisions is inadequate as a -matter of law to support a finding of no significant environmental impact. (6) Finalization of the proposed rev1s1ons to the siting regulations would not be in accord with sound agency I decisionmal<lng. Each of the above reasons alone is sufficient grounds for the Commission to _terminate the rulemaking proceeding. When taken together, the arguments are overwhelming that to proceed at this time with the siting rulemaking is contrary to sound public policy concerning the protection of the

• public health and safety and the environment from radiological hazard and disruptive of internationally accepted safety norms regarding the siting and design of nuclear power plants .

Newman & Holtzlnpr, P.C. hge11

II. MAJOR ARGUMENTS AGAINST PROPOSED CHANGES TO U.S. NUCLEAR POWER PLANT SITING REQUIREMENTS

• The International Siting Group Members urge the Commission to withdraw the proposed revisions to the siting criteria and terminate the rulemaking proceeding ..a' In the sections which follow, the basis for the ISG's request is set
• forth in detail. The arguments focus on the impact of the proposed revisions on internationally accepted standards for nuclear safety; how well the present
• regulatory framework has worked in achieving the goal of remote siting; how well the proposed revisions would contribute to reduction in nuclear power plant risk and increased defense-in-depth; and how well the proposed revisions would meet
• the Commission's stated objectives in the October 20, 1992 Federal Register notice {FAN) proposing the revisions.

I A. Adoption of the Proposed Revisions to the Commission's Demographic Regulations in 1Q CFR Part 100 Will Do Maior Damage to the Evolving International Consensus on Nuclear Safety Standards and Lead to Needless Inconsistency Between U.S. Nuclear Safety Standards tor the Siting of Nuclear Power Plants and Standards of the International Atomic Energy Agency UAEAl and National Standards in iSG Member Countries, The Existing Demographic Regulations in 10 CFR Part 100 Have Worked Well. Adoption of the Proposed Revisions Is Not Needed to Ensure Adequate Protection of the Public Health and Safety Nor to Achieve Site Isolation Through "Decoupling* of Nuclear Power Plant Siting and Design. Adoption of the Proposed I Revisions Will Not Provide a Substantial Increase In Protection or Contribute to Increased Defense-In-Depth, The Adverse Impacts of the Proposed Revisions Greatly Exceed Their Benefits . The associated draft regulatory guides, ln particular Draft Regulatory Guide DG-4003 I (Proposed Revision 2 to Regulatory Guide 4.7), should also be withdrawn. Newman & Holtzlnger, P.C. hgtt12

1. Adoption of the proposed rev1s1ons to the Commission's demographic regulations in 10 CFR Part 100 will do major damage to the evolving international consensus on nuclear

• safety standards and lead to needless inconsistency between U.S. nuclear safety standards for the siting of nuclear power plants and standards of the International Atomic Energy
• Agency (IAEA) and natlonal standards in ISG Member countries .
• The United States has had and will continue to have a major influence on nuclear power plant siting practices elsewhere in the world. The IAEA has
• established a wide-ranging program to provide its Member States with guidance on many aspects of safety associated with thermal neutron nuclear power reactors. The program has involved the preparation of many publications in the
• form of Codes of Practice and Safety Guides, many of which concern safe siting of nuclear facilities. Review of the siting documents reveals both the extensive
• participation by the United States in their development and the influence the United States has had on the substantive positions set forth in the documents.

The INSAG of the IAEA in its Safety Series No. 75-INSAG-3, "Basic Safety Principles for Nuclear Power Plants," ( 1988), sets forth basic safety principles for nuclear power plant siting: The choice of sites takes into account the results of

• investigations of local factors which could adversely affect the safety of the plant.

Sites are investigated from the radiological impact of the plant in normal operations and in accident conditions .

• Site characteristics which can influence the air, food-chain and water supply pathways are to be investigated, including physical characteristics, environmental characteristics, the use of land and water resources and the 'population distribution around the site .

Newman & Holtzlnger, P.C. Pat,e13

The site selected for a nuclear power plant is compatible with the off-site countermeasures that may be

• necessary to limit the effects of accidental releases of radioactive substances, and is expected to remain compatible with such measures.

The site selected for a nuclear power plant has a reliable long term heat sink that can remove energy generated

• in the plant after shutdown, both immediately after shutdown and over the longer term. 75-INSAG-3 at 23, 26.

An IAEA Code of Practice and a series of thirteen Safety Guides implement these I safety principles for the siting of nuclear power plants.~' The Siting Code, which was revised in 1988, establishes the objectives and basic requirements that must I be met to ensure adequate safety in the operation of nuclear power plants. The I i' The thirteen titles in the Nuclear Safety Standards series concerning siting are:

1. 50-SG-S1 (Rev. 1) - Earthquakes and Associated Topics in Relation to Nuclear Power Plant Siting (1991 );

2. 50-SG-S2 - Seismic Analysis and Testing of Nuclear Power Plants 11979);

3. 5~SG-S3 - Atmospheric Dispersion in Nuclear Power Plant Siting (1980);

4. 50-SG-S4 - Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution 11980);

5. 50-SG-S5 - External Man-Induced Events in Relation to Nuclear Power Plant Siting 11981);

6. 50-SG-S6 - Hydrological Dispersion of Radioactive Material in Relation to Nuclear I Power Plant Siting (1985);

7. 50-SG-S7 - Nuclear Power Plant Siting: Hydrogeological Aspects (1984);

8. 50-SG-S8 - Safety Aspects of the Foundations of Nuclear Power Plants (1986);

9. 50-SG-S9 - Site Survey for Nuclear Power Plants (1984);

10. 50-SG-S1 OA - Design Basis Flood for Nuclear Power Plants on River Sites (1983);

• 11 .

12. 50-SG-S1 OB - Design Basis Flood for Nuclear Power Plants on Coastal Sites (1983); 50-SG-S11A - Extreme Meteorological Events in Nuclear Power Plant Siting, Excluding Tropical Cyclones (1981 ); and

13. 50-SG-S11 B - Design Basis Tropical Cyclone for Nuclear Power Plants 11984).

• Siting experts from the NRC participated in the development of these documents, the contents of which strongly reflect U.S. siting practices.

Newman & Holtzinger, P.C. Page14

Siting Code at page 9 describes the main objective In siting nuclear power plants from the viewpoint of nuclear safety as:

• protection of the public and the environment against the radiological impact of accidental releases of radioactivity; normal radioactive releases from nuclear power plants have also to be considered. In the evaluation of the suitability of a site for a nuclear power plant the following aspects shall be considered:

(a) Effects of external events occurring in the region of the particular site (these events could be of natural or man induced origin); (b) Characteristics of the site and its environment which could influence the transfer of released radioactive material to man;

• (c) Population density and distribution and other characteristics of the external zone in relation to the possibility of implementing emergency measures and the need to evaluate the risk to individuals and the
• population .

Methods and solutions set out in the siting guides provide assurance that plants can be sited without undue risk to the health and safety of the general public. Together the Siting Code and guides establish an essential basis for safety in the siting of nuclear power plants, including the desirability of keeping reactors away from densely populated centers.!! The Siting Code and siting guides

 !!/    Safety Guide No.50-SG-S4, "Site Selection and Evaluation for Nuclear Power Plants with Respect to Population Distribution," states at 2 ( 1980):

• Countries which have developed their own nuclear power programmes from the beginning have, as far as has been practicable, begun by selecting sites in regions away from population centres and with low population densities. As experience was acquired and with technological progress, some of these countries were able to justify the choice of sites away from (continued ... )

Newman & Holtzlnger, P.C. P-,,.16

emphasize that "it is essential to ensure that all site-related characteristics have been taken into account" during the selection of the preferred candidates. ~

• IAEA Safety Guide No. 50-SG-S9, "Site Survey for Nuclear Power Plants," at 10

( 1984). That guide identifies fourteen ( 14) safety-related site characteristics to be

• evaluated during the site selection process, of which population distribution is but one)!! The guide recognizes the difficulty in comparing sites based on population and suggests that "[i]t may be appropriate to compare all other site characteristics,
• and then to evaluate the sites independently from the point of view of population distribution." kL. at 32.
• During the original development of these publications and during the revision process, care has been taken to ensure that all Member States of the
• §!( *** continued) population centres but with higher population densities. Member States embarking on a nuclear power programme may consider it prudent to give the greatest preference to sites with a low population density in the region.

The other thirteen are:

                    -- Surface faulting
                    -  Seismicity
                    -  Suitability of subsurface material Vulcanism Flooding Extreme meteorological phenomena
                    -  Man-induced events
                    -  Dispersion in air
                    -- Dispersion in water Emergency Planning Land use Availability of cooling water
                    -  Other site characteristics as appropriate, such as avalanche, landslide, surface collapse .

• IAEA Safety Guide No. 50-SG-S9 at 10-13.

Newman & Holtzinger, P.C. Pt,ge15

IAEA, in particular those with active nuclear power programs such as the United States, provide their input and that the resulting standards embody an international

• consensus. Indeed, the United States, in particular Commission representatives, have played an important role in the development of these guidance documents.

Each significantly reflects regulatory practices in the United States, as the foregoing discussion demonstrates. One of the IAEA's hopes for the revised Siting Code is that it will be used, accepted and respected by Member States as a basis

** for the regulation of the safety of power reactors within the respective national legal and regulatory frameworks.        ~     Siting Code at Foreword.

Siting, along with emergency planning and design, is viewed as an important element of defense-in-depth, a fundamental safety principle underlying the use of nuclear power. The INSAG of the IAEA describes defense-in-depth as

• follows:

                     'Defense in depth' is singled out amongst the fundamental principles since it underlies the safety technology of nuclear power. All safety activities, whether organizational, behavioral or equipment related, are subject to layers of overlapping provisions, so that if a failure should occur it would be compensated for or corrected without causing harm to individuals or the public at large. This idea of multiple levels of protection

• is the central feature of defence in depth, and it is repeatedly used in the specific safety principles that follow.

Two corollary principles of defence in depth are defined,

• namely, accident prevention and accident mitigation.

These corollary principles follow the general statement of defence in depth. Safety Series Report No. 75-INSAG-3 (1988) at 13. See also Appendix to the report, which provides an expanded discussion of

• defense-in-depth.

NBWmlln & Holtzinger, P.C. Pllge17

The NRC, in discussing comments received on its 1980 Notice Of Intent to prepare an Environmental Impact Statement (EIS) in connection with revision of the I demographic regulations In Part 100 (45 Fed. Reg. 79,820 (December 2, 1980)), stated:

• Siting, design, and emergency planning are three factors which each in its own way goes as far as is reasonable toward protecting the public health and safety. This is the 'defense-in-depth' concept. . . . [A rulemaking on any one] must consider the premises of the other two.

I NUREG-0833, 'Environmental Impact Statement on the Siting of Nuclear Power Plants Scoping Summary Report,' at 13 (December 1981 ) . The foregoing is not meant to imply that the United States is bound I to implement the codes and guides of the IAEA's Nuclear Safety Standards (NUSS) program and may not change its regulations. However, it is meant to suggest that, in the spirit of acceptance of and respect for the NUSS program, account I should be taken of the implications for change upon the international consensus standards encompassed by the NUSS program, when changes to national safety standards, sucft as 10 CFR Part 100, are being considered. It is in this spirit that ISG Members are o:ttering comments on the proposed revisions to the demographic regulations in Part 100. As discussed in I the sections which follow, the proposed revisions will do major damage to the evolving international consensus on nuclear safety standards and lead to needless

• inconsistency between* U.S. nuclear safety standards for the siting of nuclear power plants and standards of the IAEA and national standards in ISG Member countries. Specifically, adoption of the revisions by the United States would likely Newman & Ho/tzJnfler, P. C. Pat,e1B

force fundamental reconsideration of international site safety standards and the adequacy of shes selected in accordance with present standards. However, the I changes being considered by the NRC are not necessary to ensure adequate protection of the public health and safety; they will not result in improved safety; and they have the potential for destablizing an international safety framework that has worked well in the selection of suitable nuclear power plant sites in both the U.S. and elsewhere in the world. I

 **                 For example, the guidelines used by the Japan Atomic Energy Commission in reactor site evaluation closely resemble the current U.S. siting framework. An exclusion area surrounds each site, which is, in turn, surrounded by a low population zone. The reactor must also be sited away from densely populated areas. The size of the exclusion zone and population distances are

• chosen to limit radiation effects in the unlikely event of an accident. Fixed demographic limits are not imposed.

In Canada, the suitability of a site is based on the risk to the population (doses under norr:nal operation and potential doses in accidents). The risk is based on a conservative prediction of actual plant performance and site

• characteristics and is evaluated using all the exposure pathways, including land contamination, and such site characteristics as meteorology. Societal judgment about the acceptability of the risk is also a key factor in determining site suitability .
• As in Japan, fixed demographic limits are not imposed.

In France, the current practice of the safety authorities Is to pay

• special attention to demographics while assessing the suitability of potential sites Newman & Holtzlnger, P.C. Page19

for nuclear power plants. French safety authorities rely on a site-by-site analysis, which encompasses both the safety features of the plant design and the suitability I of the local background with respect to population distribution around the site, existing thoroughfares and the ability to take emergency action. Fixed

• demographic limits are not imposed.

Taiwan, Belgium and KoreaZ' base their siting practices on U.S. siting standards. Taiwan establishes exclusion area distances in accordance with the I evaluated potential radiological consequences and limited available site area, making use, in particular, of multi-reactor sites. Korea's approach is similar. I Instead of detailed Belgian-specific rules (apart from a licensing procedure, an inspection system and ICRP-type rules), Belgium safety authorities apply NRC rules in so far as practicable using a transposition process. I The proposed changes to the U.S. siting standards, if adopted, will force reconsideration of siting practices elsewhere in the world. For the additional reasons set forth in the sections which follow, the members of the International Siting Group urge the Commission to withdraw the proposed revisions to the Commission's siting regulations, along with the associated draft Regulatory Guide I DG-4003 (Proposed Revision 2 to Regulatory Guide 4. 7) . 2/ We have been able to obtain information about Belgian and Korean standards and, hence,

• are including it here, even though no Belgian or Korean organization is an ISG Member.

Newman & Holtzinger, P.C. Page20

2. The existing demographic regulations in 10 CFR Part 100 have worked well. Adoption of the proposed revisions is not needed to ensure adequate protection of the public health and safety

• nor to achieve site Isolation through "decouplingn of nuclear power plant siting and design. Adoption of the proposed revisions will not provide a substantial Increase in protection or contribute to increased defense-In-depth. The adverse impacts of the proposed revisions greatly exceed their benefits .
• a. The existing demographic regulations In 10 CFR Part 100 have worked well.

A basic assumption underlying the proposed changes to the

• demographic regulations in 10 CFR Part 100 (Part 100) is that plant design features have improved as a result of applying Part 100, but that site isolation has
• been de-emphasized. This crucial assumption -- that site isolation has been de-emphasized -- drives the proposed revisions, but has no basis in fact. As such, it does not provide a valid basis for the demographic criteria in the proposed rule.

I Contrary to this assumption, under the existing NRC regulatory framework for nuclear power plant siting, based on 10 CFR Parts 50, 51 and 1 00 and on the philosophy and guidelines published in Regulatory Guide 4. 7 {Rev. 1), U.S. nuclear power plants have been sited away from highly populated areas and the Commission's remote siting objective has been achieved. Very recently, NRC

• Staff representatives underscored this fact when meeting with members of the international community in January 1993, attending a meeting of the Committee on Nuclear Regulatory Activities (CNAA) of the OECD in Paris. The Staff's briefing
• charts stated:

Use of Reg. Guide 4. 7 [Rev. 1] in conjunction with Part 100 provides effective means to keep reactors away

• from densely populated centers. NRC Staff Trip Report Newman & Holtzins,er, P.C. hgt,21

Concerning CNRA Meeting, enclosure at 8 (February 5, 1993) .

• I. Summary of existing NRC framework for siting nuclear power plants.

The demographic regulations in Part 100 were promulgated in 1962 .

• The Supplementary Information for these safety regulations identified as a basic objective of the regulations assurance that "the cumulative exposure dose to large numbers of people as a consequence of any nuclear accident should be low in
• comparison with what might be considered reasonable for total population dose .

 . . . [Another objective Is to] provide for protection against excessive exposure

• doses to people in large centers, where effective protective measures might not be feasible." 27 Fed. Reg. 3,509 (April 12, 1962). The regulations identified site evaluation factors, which included population density and "use characteristics"
• ti.JL., characteristic human activities) of the site environs; and provided guidance for determining the suitability of the proposed site on the basis of a dose assessment which took into account:

the characteristics of the reactor design; population density and use characteristics, including the

• exclusion area, low population zone and population zone distance; and physical characteristics of the site, including seismology, meteorology, geology, and hydrology .
• Flexibility of application was an important aspect of implementation to make certain that the concept of environmental isolation did not receive undue emphasis I

hge22

I and to recognize the importance of engineered safeguards In meeting the regulation's objective. See id... I Appendix A to 10 CFR Part 50 establishes the minimum requirements I for the principal design criteria for nuclear power plants. A number of these criteria are directly related to site characteristics, as well as to events and conditions outside the nuclear power unit. Part 50 also specifies emergency I planning and preparedness requirements. Compliance with the National Environmental Policy Act of 1969 I (NEPA)!' is also a factor in nuclear power plant siting. NEPA requires that a cost benefit analysis be completed before any major federal action significantly affecting the human environment is undertaken. Nuclear power plant siting, being the initial I step of a major federal action to license a plant for operation, is necessarily encompassed by the provisions of NEPA)!.! NRC requires preparation of alternative site studies which balance environmental costs and benefits of several preselected sites. Population characteristics are among the site characteristics entering into the cost-benefit balancing. The site selected by the applicant is to I be among the best reasonably to be found for which no obviously superior alternative has been identified.w I fl/ 42 U.S.C. § 4332(2}(c) (1988). Calvert Cliffs Coordinating Committee v. AEC. 449 F.2d 1109. 1112 (D.C. Cir. 1971 ).

          ~   10 CFR Part 51 and Regulatory Guide 4.2, Rev. 2, "Preparation of Environmental I         Reports for Nuclear Power Stations" (1976).

Newman & Holtzinger, P.C. Page23

I Regulatory Guide 4. 7 (Rev. 1), *General Site Suitability Criteria for Nuclear Power Stations," was issued In November 1975. 111 The guide is I intended to assist applicants in the initial stage of selecting potential sites for nuclear power stations. Regulatory Gulde 4. 7 provides guidance related to both I the site's safety and Its environmental qualities. Regulatory Gulde 4. 7 implements the safety criteria in 10 CFR

  §  100. 11 pertaining to demographics as follows.             Each nuclear power plant I                                          J applicant must determine the following:

1. An exclusion area of such size that an individual located at any point on its boundary for two hours immediately I following onset of the postulated fission product release would not receive a total radiation dose to the y,.,hole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from Iodine exposure.

I 2. A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

3. A population center distance of at least one and one-third times the distance from the reactor to the outer I boundary of the low population zone. In applying this guide, the boundary of the population center shall be

                  -determined upon consideration of population distribution. Political boundaries are not controlling in the application of this guide. Where very large cities are I

In November 1992, the NRC issued for public comment Draft Regulatory Guide DG-4003 (Proposed Revision 2 to Regulatory Guide 4.7). Relevant differences between Rev. 1 and Proposed Rev. 2 are noted where appropriate. The changes in Proposed Revision 2 I conform to the proposed changes to 10 CFR Part 100. Newman & Holtzinger, P.C. I

Involved a greater distance may be necessary because of total Integrated population dose consideration. 121

• Regulatory Guide 4. 7 (Rev. 1) also provides guidance concerning consideration of alternative sites when numerical demographic criteria given in the guide are exceeded. The" guide states:
• If the population density, including weighted* transient population, projected at the time of initial operation of the nuclear power station exceeds 600 persons per square mile averaged over any radial distance out to 30
• miles (cumulative population at a distance divided by an area at that distance), or the projected population density over the lifetime of the facility exceeds 1000 persons per square mile averaged over any radial distance out to 30 miles, special attention should be
• given to the consideration of alternative sites with lower population densities (emphasis added).~'

In sum, Regulatory Guide 4. 7 (Rev. 1) makes clear that "[t]he decision

• that a station may be built on a specific candidate site is based on a detailed evaluation of the proposed site-plant combination and a cost-benefit analysis 10 CFR 100.11 (1993) .

                   § ll'    Regulatory Guide 4. 7 (Rev. 1l at 16. Draft Regulatory Guide DG-4003 states substantially the same at 9:

• As set forth in 10 CFR Part 100, nuclear power station sites should be located In areas with low population density. If the population density of a proposed site ( 1 ) exceeds 500 people per square mile averaged over any radial distance out to 30 miles or (2) is projected to exceed 1000 people per square mile averaged over any radial distance out to 30 miles (50 kilometers) 40 years after the time of site approval, the applicants should give special attention to I alternative sites.

Newman & Holtzlnger, P.C. hge26

comparing it with alternative site-plant combinations . . . " (emphasis added) .w As discussed above, the analysis includes consideration of site safety issues and

• environmental issues. The site safety issues include geologic/seismic, hydrologic and atmospheric characteristics of proposed sites; potential effects on the station
• from accidents associated with nearby industrial, transportation, and military facilities; and population densities in the site environs as they relate to protecting the general public from the potential radiation hazards of postulated serious
• accidents ..!!' When an applicant's preferred site does not meet the numerical population density guidelines in Regulatory Guide 4. 7 (Rev. 1), consideration is
• given to alternative sites .

Like the international site safety standards described above, implementation of the Commission's current site safety regulations requires a

• balanced account of all factors contributing to safety and to reduced risk of 14/

Regulatory Guide 4. 7 (Rev. 1I at 1 . While Draft Guide DG-4003 replaces "site-plant combination" with "site," it makes clear at 3 that *cslite selection involves considerations of public health and safety, englneerjng and design, economics, institutional requirements, environmental impacts, and other factors.* (Emphasis added.) ll/ !9.,_ at 1-2. Draft Guide DG-4003 states at 4: I Generally, the most restrictive safety-related site characteristics considered in determining the suitability of a site are surface faulting, potential ground motion and foundation conditions (including liquefaction, subsidence, and landslide potential), and seismically induced floods.

• Of atmospheric extremes, the Draft Guide states at 5:

mhe atmospheric extremes that may occur at a site are not normally critical in determining the suitability of a site because safety-related structures. systems. and components can be designed to withstand most atmospheric extremes. {Emphasis

• added.)

N.wman & Holalnger, P.C. Pt,ge 26

I accident consequences. The emphasis on dose calculations to determine exclusion areas, low population zones and population center distances places significant I importance on engineered safety features found at plants that reduce and contain potential accidental releases of radioactivity from the plant. Additionally, emphasis I is placed on emergency preparedness in that the exclusion areas surrounding the plant must be totally controlled by the reactor licensee and the low population zone immediately surrounding the exclusion area must be such that the population I number and distribution provide a reasonable probability that appropriate measures could be taken in the event of a serious accident. The guidance on Regulatory Guide 4. 7 also emphasizes that when a site is surrounded by more than 500 I persons per square mile over a radial distance 30 miles from the plant, consideration should be given to better alternative sites. I II. Existing siting regulations have achieved site Isolation. In its publication, "Demographic Statistics Pertaining to Nuclear Power Reactor Sites" (NUREG-:0348), issued in October 1979, the NRC examined in detail the demographic characteristics surrounding plant sites. NUREG-0348 discussed the results of a trend analysis and provided a variety of data on population I densities and distances from population centers. The purpose of the NUREG-0348 analysis was to determine If a trend existed toward greater site Isolation. The

• analysis performed indicated that nuclear power plants wer*e being sited away from high population areas.~'

J.!I NUREG-0348, Fig. 17. I Newman & Holtzinger, P.C. Page27

Even more significant than this result is what the data indicate about site isolation. The data demonstrate conclusively that the goal of remote siting

• was achieved in the 1970s. Of the 58 new sites docketed at the NRC from 1971 to 1979, Perryman was the only new site proposed that did not meet tt;,e
• population density requirements at 30 miles. See kl:. at Table 1 . When the staff received the application for an early site review for Perryman, it concluded that an obviously superior alternate site was preferable. The application for the Perryman
• site was subsequently withdrawn. 171 In addition, only about seven of the sites in the United States today (approximately 10%) have population densities *in
• excess of the Regulatory Guide 4. 7 (Rev. 1) guidelines. In each case, the sites comprising the seven were given construction permits prior to the adoption of the Regulatory Guide 4. 7 (Rev. 1) parameters in November 1975.w Thus, the I Commission's framework for nuclear power plant siting developed in the 1970s has achieved site isolation.

The basic assumption underlying the proposed changes to the demographic regulations in 10 CFR Part 100 -- namely, that as a result of applying Part 100, site isolation had been de-emphasized -- is taken from the August 1979 NRC "Report of the Siting Policy Task Force" (NUREG-0625))..!/ As the w !d.. at 20.

• 19{

NRC Staff Trip Report Concerning CNRA Meeting, enclosure at 7 (February 5, 1993) . NUREG-0625 was the result of several years study of possible revisions to the Commission's siting regulations. The purpose of the report was to obtain an overview of the siting policy and practice that resulted from 26 years of licensing of civilian nuclear power plants. Another purpose was to determine whether current siting policy and

• (continued ... )

Newman & Holtzlngt,,, P. C. Pa{/tl28

foregoing discussion of the NUREG-0348 data demonstrates, this critical assumption that site isolation has been de-emphasized has no basis in fact. As

• such, it cannot provide a valid basis for the proposed demographic regulation.

The lack of factual basis was recognized at the time NUREG-0625

• was undergoing internal NRC review prior to its release. Robert B. Minogue, the Director of NRC's Office of Standards Development, after reviewing NUREG-0625, stated in a letter to Daniel R. Muller, Chairman of the Siting Policy Task Force,
• dated August 15, 1979:

The implication in the discussion of past practices that the demographic features of population and distances have been getting progressively worse at licensed sites I Is not true. Indian Point, San Onofre, and Zion sites were reviewed and approved more than 10 years ago. Demographic features of current licensed sites have actually been improving somewhat since the above listed sites were approved.jQ'

• Mr. Minogue stated further, n *** we are concerned about the prospect that the report may be forced to be used as a basis for Immediate rulemaking and is inadequate for that purpose. nllf

  !!!/( *** continued)

' practice should be changed. The report made nine recommendations, the most important of which were to divorce from the siting framework the use of plant design features to compensate for unfavorable site characteristics and to develop population density and distribution limits beyond the exclusion area which would be functions of the average population in different regions of the United States. Oependance on the average population of a region meant that areas of the United States having lower average population densities might be subject to higher population density and distribution limits than more densely populated regions of the country, such as the Northeast. Z9! NUREG-0625 at 78. ill kl,. at 77. NflWrl'IMI & Holtzinger, P.C. t

Another NRC official also discussed limitations in the use of the NUREG-0625 recommendations. In a letter to Mr. Muller, dated August 14, 1979,

• Norman M. Haller, then NRC' s Director of the Office of Management and Program Analysis, recommended publication of a definitive value-impact analysis of all Task
• Force recommendations contained in NUREG-0625 before any recommendations were released for public comment. Mr. Haller stated:

                  ... the net or true cost cannot be estimated unless the

• next best alternative (namely, allowing trade-offs between distance and unique design features to be made) is also analyzed.
• In general, adoption of the distance-related recommendations in this report would appear to undermine the philosophy that reactors can operate

                . safely primarily because their designs satisfy NRC

• regulations. And, we believe that adoption of these recommendations would leave the Commission open to the charge that some existing reactors aren't safe enough (since they rely on design features) ..W In sum, no basis exists for the statement in NUREG-0625 that nuclear power plant site isolation in the United States has been de-emphasized. To the contrary, under the existing siting regulations in 10 CFR Parts 50, 51 and 100 and implementing guidance, the objective of remote siting has been achieved .

221 k!:. at 75. Newman & Holtzinger, P. C. Ptlg* 30 I

b. Adoption of the proposed revisions Is not needed to ensure adequate protection of the publlc health and safety .

• The Supplementary Information for the proposed revisions to Part 100 indicates that since promulgation of the reactor site regulations in 1962, the
• Commission has approved more than 75 sites for nuclear power reactors. 57 Fed.

Reg. at 47,803. These approvals required an affirmative finding of adequate protection of the public health and safety. We may conclude, therefore, that

 ** adoption of the proposed revisions is not necessary to ensure adequate protection of the public health and safety.
                     'Other considerations buttress this conclusion.         As recently as December 1988, the NRC denied a petition for rulemaking (PRM) to amend Part 100 to specify demographic criteria to be met for all new nuclear power plant I

sites. 53 Fed. Reg. 50,232 (December 14, 1988). The NRC denied the petition for the following reasons: ( 1) it would unnecessarily restrict NRC' s regulatory siting policies and procedures by elevating population density criteria above other siting criteria such as environmental and ecological factors, and (2) it would not result in a substantial increase in overall protection of the public health and safety, as compared to the ' current siting criteria when combined with calculations of potential health effects. The NRC has carefully considered the issues raised in the petition, and has taken them into account in reaching a decision on the areas which fall within its jurisdiction. lg_.. I The petitioners had requested that the Commission amend its regulations in 10 CFR Part 100 to set numerical limits on allowable population t density around nuclear power reactor sites. The amendments to 10 CFR § Newmt1n & Holtzlnger, P.C. Pl,ge31 I

100. 11 (a) proposed by the petitioners would set 0.4 miles and 3 miles as the minimum distances for the outer boundaries of the exclusion area and the low

• population zone, respectively. A new section of 10 CFR' § 100. 12 proposed by the petitioners would set a maximum population density of 400 persons per square
• mile averaged over any radial distance out to a distance of 40 miles. The petitioners proposed that the Commission also deny Construction Permit applications where, during the effective period of the plant's license, the maximum projected population density would exceed 800 persons per square mile averaged over any radial distance out to a distance of "!-0 miles. Additionally, the petitioners I proposed that all population figures and projections include transit populations.

The Commission amplified its reasons for denial as follows: At first glance, it might appear that the NRC's

• population density siting parameters and the population density siting parameters indicated by the petitioner are similar -- 500 vs. 400 per square miles averaged over any radial distance of 40 vs. 30 miles for the initial operation of the nuclear power plants. However, the real difference between the NRC's and the petitioner's population density siting requirements is regulatory flexibility. The NRC's siting requirements allow for the consideration of alternative sites with superior environmental parameters, e.g., suitable meteorological, natural resources and water temperature conditions or
• superior geophysical conditions, e.g., suitable geologic, hydrologic, and tectonic conditions if the population density parameters cannot be met. However, on the other hand, the petition's siting requirements would automatically eliminate any site from further I consideration if specific population density criteria are not met regardless of any_ mitigating factors.

The NRC believes that Regulatory Guide 4. 7 [Rev. 1J adequately addresses population density siting considera-tions and that no new rulemaking as proposed Newman & Holtzlnpr, P.C. P,,ge32

by the petitioners is justified at this time. Also, the petitioner offers no basis for the specific numerical population density limits indicated in the petition .

• Therefore, the petition would not result in a substantial increase in the overall protection of the public health and safety, as compared to the current NBC siting criteria when combined with ca!culatiqns of potential health effects. kh at 50,233 (emphasis added).

I Consideration of Commission findings concerning the residual risk from severe accidents and compliance with the Commission's safety goals also I testifies to the adequacy *of the present siting regulations in ensuring adequate protection of the public health and safety. In 1985, the Commission issued its "Policy Statement on Severe Reactor Accidents Regarding Future Designs and I Existing Plants." 50 Fed. Reg. 32,138 (August 8, 1985). The Commission emphasized that "[o]n the basis of currently available information, the Commission I concludes that existing plants pose no undue risk to public health and safety and sees no present basis for immediate action on generic rulemaking or other regulatory changes for those plants because of severe accident risk." kl.. The Commission stated that its "severe accident policy is that the Commission intends to take all reasonable steps to reduce the chances of occurrence of a severe accident involving substantial damage to the reactor core and to mitigate the consequences of such an accident should one occur." kt.:. at 32,139. In promulgating the Safety Goal Policy Statement, the Commission stated its belief that "Current regulatory practices are believed to ensure that the basic statutory requirement, adequate protection of the public is met." 51 Fed. Reg. 30,028, at 30,029 (August 21, 1986). The Safety Goal Policy Statement "expresses the Newman & Holtzfnger, P. C. i'age33

Commission's views on the level of risks to public health and safety that the industry should strive for in its nuclear power plants" and provides a framework

• "for testing the adequacy of and need for current and proposed regulatory requirements" in order to "lead to a more coherent and consistent regulation of
• nuclear power plants, a more predictable regulatory process, a public understanding of the regulatory criteria that the NRC applies, and public confidence in the safety of operating plants." kL.
• In 1990, NUREG-1150, "Severe Accident Risk Assessment for Five U.S. Nuclear Power Plants," was published; and in 1992, NUREG-1465, "Accident
• Source Terms for Light Water Nuclear Power Plants," was issued for public comment. Using state-of-the-art risk assessment techniques, NUREG-1150 studied the risks from severe accidents in five nuclear power plants representative I

of plants presently in operation today in the United States. Draft NUREG-1465, using updated knowledge about severe LWR accidents, !Ind the resulting behavior of the released fission products, developed over a 30-year period, provides a postulated fission product source term released into containment. The issuance of the Severe Accident and Safety Goal Policy I statements and the enhanced capability to evaluate severe accident risk and understand severe accident source terms, as demonstrated by NUREG-1150 and NU REG- 1465 results, allow the Commission to evaluate the risk significance of any I revisions to the Part 100 regulations that might be proposed. Indeed, the purpose of the Commission's earlier suspension of the siting rulemaking was, first, to I develop such a capability and, then, to utilize that capability in evaluating the Newm*n & Holtzlnger, P. C. hge34

safety significance of alternative proposals to revise the regulatory framework for siting nuclear power plants.W

• In the 1992 Supplementary Information for the proposed changes to the siting regulations, the Commission relied on NUREG-1150 as one of:
• [N]umerous risk studies on radioactive material releases to the environment under severe accident conditions

[which] have all confirmed that the present siting practice is expected to effectively limit risk to the public. 57 Fed. Reg. at 47,803 . Figure 13.2 of NUREG-1150, reproduced on the next page, demonstrates that the quantitative health effects objectives specified in the Safety Goal Policy Statement

• In the Summer of 1980, the Commission began an effort to revise the siting criteria in 10 CFR Part 100. On July 29, 1980, the NRC issued an Advance Notice of Proposed Rulemaking (ANPR) (45 Fed. Reg. 50,350), in which the Commission announced its intention to revise the reactor siting criteria and requested comments on seven of the nine recommendations of the Siting Policy Task Force, as well as certain alternative approaches.

In conjunction with the rulemaking effort, the Commission also issued a Notice of Intent {NOi) to prepare an Environmental Impact Statement (EIS) (45 Fed. Reg. 79,280 (December 2, 1980)). The NOi, among other things, identified the technical approach to detailed analyses that would be followed in developing the bases for any proposed revisions. ~ m.:. at 79,822-23. In December 1981, the NRC published the Scoping Summary Report for the EIS (NUREG-0833). The report addressed comments received on both the ANPR and the NOi and provided further discussion of the efforts which would be undertaken to develop an adequate technical basis for any revisions. The report recognized that the siting rulemaklng must take into account premises concerning reactor design and emergency planning. ~ NUREG~0833 at 13. It also restated that "a systematic evaluation of accident consequences for a full range of reactor accidents would be a fundamental part I of the technical basis for the siting rulemaklng *..*

• hi,. at 20. See also id... at 14 regarding consideration of a full range of accidents in establishing siting criteria.

Shortly thereafter, the Commission directed the NRC Staff to suspend work on revision of the siting {demographic) criteria until safety goals were developed and a reassessment of source tenns was completed. ~ NUREG-0885, "U.S. Nuclear Regulatory Commission t Policy and Planning Guidance* (January 1982). Page36

13. Resource Document I

Individual early fatallty/ry 1.0E-06 _ _ _ _ _.....:__ _....:;._,:~-------.---,r----L-eg_e_nd~-, Safety Goal 1.0E-07 I 1.0E-08 I 1.0E-09

                                              ~

1.0E-10

• ~

 -          1.0E-11 SURRY           PEACH BOTTOM SEQUOYAH Individual latent cancer fatallty/ry 1.0E-05 =-----------~__;_------,---L-eg_e_nd_--,

GRAND GULF ZION G611

                         ~Safety Goal 1.0E-06                                                                      lflean median

• 1.0E-07

                                                                                    ~  tiS 1.0E-08 I

SURRY lt 1.0E-10 L-_J..._ _ _ _.1..__ _ _---1_ _ _ _--1-_ _ _ _...,___ PEACH SEQUOYAH GRAND ZION BOTTOM GULF Note: As discussed in Reference 13.23, estimated risks at or below 1E-7 per reactor year I should be viewed with caution because of the potential impact of events not studied in the risk analyses. Figure 13.2 Comparison of individual early and latent cancer fatality risks at all plants (internal initiators). I NUREG-1150

I have been metw and, consequently, that the basic statutory requirement of adequate protection has been met under present (site-plant combination} siting I practice. Of particular interest are the results from one of these five plants, the I Zion plant..W The Zion population density figure is close to the 500 people/square mile value in the proposed rule. As Figure 13.2 shows, the Zion I I I The prompt fatality health effects objective, a measure of whether the individual safety goal has been met, states: The risk to an average individual in the vicinity of a nuclear power plant of prompt fatalities that might result from reactor accidents should not exceed one-tenth of one percent (0.1 percent) of the sum of prompt fatality risks resulting from other accidents to which members of the U.S. population are generally exposed. 51 Fed. Reg. at 30,030. I The latent cancer mortality health objective, a measure of whether the societal safety goal has been met, states: The risk to the population in the area of a nuclear power plant of cancer fatalities that might result from nuclear power plant

• operation should not exceed one-tenth of one percent (0.1 percent) of the sum of cancer fatality risks resulting from all causes. H.!:.

The average population density around the, Zion plant was 457 people/square mile in the early 1980s, when calculated over an area with a 30 mile radius originating at the plant site. NUREG/CR-2239, *rechnical Guidance for Siting Criteria Development,* (November

• 1982).

NIIWmMI & Holtzlng<<, P. C.

I early and latent fatality risks are quite low and well below the Safety Goals, thus supporting the conclusion that present siting practice has worked well in limiting I public risk from nuclear power plant operation. Future plants, with even greater safety capabilities, would show even greater margins. The Severe Accident Policy I Statement requires, in effect, that all new plant designs meet certain "fundamental criteria" to reduce severe accident risk and that cost-effective features of a preventive or mitigative nature be included in the design. The Safety Goal Policy I Statement provides guidance in determining cost-effectiveness. Taken together, these two policy statements ensure adequate protection of the public health and I safety and a very low level of residual risk. In sum, adoption of the proposed revisions is not necessary to ensure adequate protection of the public health and safety .

• c. Adoption of the proposed revisions is not needed to achieve site Isolation through *decoupling" of nuclear power plant siting and design.

In .addition to specifying fixed, numerical requirements in Part 100, the proposed revisions relocate certain design requirements pertaining to plant design to 10 CFR Part 50, "thereby effectively decoupling siting from plant I design." ~ Supplementary Information for the proposed revisions at 57 Fed. Reg. 47,803. The proposed decoupling is directed at ensuring that engineering safeguards or advanced safety features are not used as a substitute for site

'  isolation. The proposed decoupling is intended to implement the Siting Policy Task Force goal:

t Newman & Holtzlnpr, P.C.

I To strengthen siting as a factor in defense in depth by establishing requirements for site approval that are independent of plant design consideration. The present I policy of permitting plant design features to compensate for unfavorable site characteristics has resulted In improved designs but has tended to deemphasize sjte isolation. NUREG-0625 at iii (emphasis added). I As discussed in the preceding sections, the Commission's present siting requirements in Parts 50, 51 and 100, and implementing guidance in Regulatory Guide 4. 7, have achieved site isolation. In particular, consideration of I alternative sites when numerical demographic criteria in Regulatory Guide 4. 7 have been exceeded ensures site isolation. Thus, decoupling is not needed to achieve I site isolation. The fact that the proposed revisions to the Commission's demographic regulations in Part 100 are unnecessary to achieve isolation in nuclear I power plant siting by decoupling siting and design is buttressed by consideration of the provisions of 10 CFR Part 52, "Early Site Permits; Standard Design Certifications; and Combined Licenses for Nuclear Power Plants," adopted by the Commission in 1989. With the adoption of Part 52, it is not necessary to revise Part 100 to accomplish such "decoupling." Part 52 promotes the use of pre-approved standardized designs which are certified (approved) by the Commission in a design certification rulemaking proceeding, conducted pursuant to Subpart B of Part 52 ..w Certification does not involve consideration of specific sites so I Subpart B to Part 52, "Standard Design Certifications," sets forth the requirements and procedures for "certification,* or pre-approval through a hybrid rulemaklng process, of new standardized nuclear power plant designs to ensure that all safety-related design issues are resolved prior to the purchase or construction of a new standard plant .

*                                                                                      (continued ... )

N.wm.n & Holtzinger, P.C. hge3B

I there is no tailoring of a design to compensate for site deficiencies. Certified designs incorporated by reference into a nuclear power plant application would not I be subject to further review or challenge in a licensing proceeding unless the applicant proposed to make changes to the certified design. In particular, Subpart I C of Part 52 provides for issuance of a combined construction permit/operating license and permits applicants for such to incorporate by reference certified designs into the application. These limitations on review and challenge in a nuclear I power plant licensing proceeding implicitly encourage the selection of sites falling within the siting envelope specified in the design certification and discourage the selection of sites which would require changes to the certified design. I Additionally, Part 52 provides for early approval of sites in a licensing proceeding conducted in accordance with Subpart A of Part 52. 271 While general

• 261

( *** continued) Specifically, an applicant for design certification must provide, among other things, all technical information which is required of applicants for construction permits and operating licenses by 10 CFR Parts 20, 50 and its appendices, 73 and 100, which is technically relevant to* the design and nm site-specific. Subpart B further requires the design certification applicant to prepare a design-specific probabilistic risk assessment ( 10 CFR § 52.47(a)(1 ){v)), and include proposed inspections, test, analyses and acceptance criteria which are necessary and sufficient to provide reasonable assurance that if the tests, inspections and analyses are performed and the acceptance criteria met, a plant which I references the design is built and will operate in accordance with the design certification.

               ~ 10 CFR § 52.47(a)(1)(vi) (1993).

111 Subpart A of Part 52, "Early Site Permits," allows for early resolution of site-related safety and environmental issues and authorizes pre-approval of sites for new nuclear power plants

               - separate and apart from design approval. Subpart A will allow utilities to "bank" sites

• for new nuclear power plant facilities before the need for them has materialized and independent of design details for a nuclear power plant tailored to the site. In particular, an applicant for an early site permit must provide, among other things, a description of:

(1 l the number, type, and thennal power level of the facilities for which the site may be used; (2) the anticipated maximum levels of radiological and thermal effluents each facility will produce; (3) the type of cooling systems that may be associated with each facility; (4) I (continued ... ) Ntwm11111 & Holtzinger, P.C. Page39 I

I design information must be specified In an early site approval application, Part 52, when viewed in its entirety, creates incentives to choose sites suitable for use as

• the location of a plant of certified design.

ISG Members believe that there are significant disadvantages to

• decoupling siting and plant design so that siting does not take into account plant design features. First, decoupling siting and plant design in the manner proposed is not necessary to achieve site isolation. Second, decoupling eliminates an
• accepted measure of the overall merit of the site-plant combination with respect to safety without providing for a substitute. Third, some of the major advances in design have come because designers wished to ensure greater safety for a

  !?!( *.* continued) the seismic, meteorological, hydrologic, and geologic characteristics of the proposed site

• (as set forth in existing Appendix A to 10 CFR Part 100); and (5) the existing and projected future population profile of the area surrounding the site. ~ 10 CFR § 52.17{a)(1 HiHviii)

(1993). Emergency planning must also be considered at the ear1y site permit stage. Three options are available to the applicant ranging from Identification of significant impediments to a complete integrated plan. At a minimum, an early site permit applicant must identify any significant impediments to emergency planning and list the contacts and arrangements made with state, local and federal agencies with emergency planning responsibilities. Such impediments, if any, will be assessed in determining whether any alternative site is obviously superior. ,S,fil! 10 CFR § 52.17(b) {1993). In addition, an applicant may either request NRC approval for major features of an emergency plan, or approval of a complete

• integrated emergency plan. ~ 10 CFR § 52.17(b){2) {1993).

Moreover, an early site permit applicant must provide a complete environmental report as required under Part 51, which includes an evaluation of atternative sites to determine whether any "obviously superior attemative* to the proposed site exists . .SU 10 CFR § 52.17{a)(2} (1993). The Part 52 early site permit process does, however, require certain

• limited consideration of design features. Specifically, an applicant must provide a description and safety assessment of the proposed site, *with appropriate attention to features affecting facility design." 10 CFR § 50.34(a)(1) (1993); .IM~ 10 CFR

              § 52.17(a)(1) (1993). Such an assessment must *contain an analysis and evaluation of the major structures, systems and components of the facility which bear significantly on the acceptability of the site under the site evaluation factors identified in Part 100.... "

• J..g_,_

Nt!IWmlln & Holtzinpr, P.C. hge40

I specific site. For example, the multi-unit vacuum containment for plants of CAN DU design originated as *a means to ensure additional safety for the Pickering

• plant because of its location. The same containment concept was later used for multi-unit stations at remote sites, even though the additional safety was not
• needed to ensure adequate protection of the public health and safety. More generally, in countries which have very few acceptable sites due to basic land availability or cooling water supply or public acceptance, the regulatory structure should encourage designers to innovate to reduce public risk in making use of the available sites. The restrictions on siting contemplated by the proposed revisions to the demographic requirements in Part 100 will have to be ignored by such countries (with consequent justification as to why imposition of NRC-like requirements is not necessary) or will pose real and unnecessary restrictions to the growth of nuclear power in places where it is most needed.

In sum, adoption of the proposed revisions to the Commission's site safety regulati~ns in Part 100 is not necessary to ensure site isolation through decoupling of nuclear power plant siting and design. As discussed above, the regulatory position stated in Regulatory Guide 4. 7 to consider alternative sites when numerical population density criteria are exceeded provides an appropriate context for ensuring site isolation. Newman & Holtzinpr, P.C.

d. Adoption of the proposed revisions to the demographic regulatf ons in Part 100 will not provide a substantial increase in protection nor contribute to Increased

• defense-In-depth. The adverse impacts of the proposed revisions greatly exceed their benefits.

i. NRC decisional framework when action Is not necessary to ensure adequate protection .

• Discussion in the previous sections established that no change to the Commission's present demographic regulations in Part 100 is necessary to ensure
• adequate protection of the public health and safety. As discussed above, one of
• the primary reasons the Commission adopted the Safety Goals was to establish a coherent and consistent set of safety regulations and provide a means to determine whether future safety regulations were necessary. Consistent with the Atomic Energy Act, safety requirements must be imposed when they are necessary
• to provide adequate protection of the public health and safety.

requirements may be imposed when they are not needed for adequate protection, Safety if they are cost-effective and afford a substantial Increase in protection. When it can be shown that the Safety Goals are met, the Commission has indicated that an increase in protection cannot be substantial and that no additional safety requirements need or should be imposed. By setting limits on population density in Part 100, adoption of the proposed rule would establish additional safety requirements beyond the Safety Goals which are not needed and of little benefit, thus creating an inconsistency in the Commission's regulatory philosophy. Consistent with the decisional framework for implementation of the Safety Goal Policy Statement, the Commission, in deciding whether to adopt the Newman & Holtzlnger, P.C. t

proposed changes to the demographic regulations in Part 100, should consider (1) whether they will provide a substantial increase in protection of the public

• health and safety and (2) whether the benefits from the changes will outweigh the impacts. Absent an affirmative finding on both of these questions, the

' Commission should not adopt the proposed changes. As discussed below, ISG Members believe that adoption will not provide a substantial increase in protection and that the impacts will far outweigh the benefits.

• In its safety regulation of nuclear power plants, the Commission distinguishes between changes necessary to ensure adequate protection of the public health and safety and changes imposed to effect safety improvements beyond the minimum needed for adequate protection. This principle was clarified as a result of litigation about the initial formulation of the so-called nbackfit rule" (10 CFR § 50.109), which distinguishes between the two kinds of changes ..a!' When establishing safety requirements which are not necessary for adequate protection of the public health and safety:

[Tihe Commission shall require the backfitting of a facility only when it determines . . . that there is a substantial increase In the overall protection of the public health and safety or the common defense and t security to be derived from the backfit and that the direct and Indirect costs of implementation for that facility are Justified in view of this increased protection. 10 CFR § 50.109(a)(3) (1993) (emphasis added). W ISG Members believe that the direction the Commission has provided on the application of the Safety Goal Policy to backfit decisions is relevant and should guide Commission decisionmaking concerning adoption of the proposed revisions to Part 100 to ensure regulatory coherence and consistency in establishing Commission requirements. Newman & Holtzlnger, P. C.

In adopting the present version of the backfit rule, the Commission stated:

• In this rulemaking the Commission has adhered to the following safety principle for all of its backfitting decisions. The Atomic Energy Act commands the Commission to ensure that nuclear power plant operation provides adequate protection to the health and safety of the public. In defining, redefining or enforcing
• this statutory standard of adequate protection, the Commission will not consider economic costs.

However, adequate protection is not absolute protection or zero risk. Hence safety improvements beyond the minimum needed for adequate protection are possible .

• The Commission is empowered under section 161 of the I
• Act to impose additional safety requirements not needed for adequate protection and to consider economic costs in doing so. 53 Fed. Reg. 20,603, at 20,604 (June 6, 1988).

In a Staff Requirements Memorandum, "SECY-89-102 Implementation of the Safety Goals," dated June 15, 1990, the Commission It addressed the meaning of "substantial increase in protection as an application of the Safety Goals." The Commission indicated that once it could be established that the Safety ~oals have been met, any further increase in protection would not be substantial. This development allows the Commission to make decisions about proposed regulatory actions based on their safety significance .

• In sum, when examined against the backdrop of these developments, it is clear that the Commission may take into account the adverse impacts (costs) of adopting a new safety regulation, when adoption of the regulation is not I

necessary to provide adequate protection. Moreover, ISG Members believe that the Commission's decisional framework for Safety Goal implementation should be used when considering whether a proposed change in safety requirements will Newman & Holtzlnpr, P.C. Pat,e44

contribute reasonably to Increased defense-In-depth. Since imposition of fixed, numerical demographic 'criteria through changes to the present demographic

• regulations in Part 100 clearly is not necessary for adequate protection, as discussed above, it is equally clear that the adverse impacts (costs) that result
• from such imposition may be weighed against the benefits of their imposition .

When such an approach is taken to evaluate the proposed revisions to the demographic regulations in Part 100, two conclusions inexorably follow .

• First, adoption of the proposed revisions will not result in a substantial increase in protection nor contribute to increased defense-in-depth. Second, the adverse
• impacts from their adoption will far outweigh the benefits. Based on such conclusions, the ISG Members urge the Commission to withdraw the proposed revisions and terminate the present rulemaklng proceeding.

I ii. No substantial Increase in protection or defense" in--depth is afforded by the proposed revisions. Concerning the first point, in 1988 the NRC denied a 1976 petition for rulemaking {PRM-100-2) to set more restrictive siting distances and population densities than in Regulatory Guide 4. 7 (Rev. 1), partly on the grounds that granting of the petition would not result in a substantial increase in the overall protection

• of the public health and safety. {Sufilllllil Section 11.A.2.b.) Clearly then, the less restrictive numerical criteria of Regulatory Guide 4. 7 (Rev. 1) being proposed for
• Inclusion in the demographic regulations could not provide a substantial increase in protection through the mere act of codification.

I Newman & Ho/tzinpr, P.C. Page46

Similarly, the mere act of co~ification will not contribute to increased defense-in-depth. The Commission and Staff have made It abundantly clear that

• the proposed changes will have an insignificant impact on risk. Specifically, the Supplementary Information for the proposed rule makes clear that present practice
• has effectively limited risk to the public. ~ supra Section 11.A.2.b, concerning the discussion of NUREG-1150 results.) In discussions before the Commission's Advisory Committee on Reactor Safeguards (ACRS) In January 1992, the NRC
• Staff tried to explain the basis for the proposed revisions, as well as provide some background on the thinking of the 1979 Siting Policy Task Force. Mr. Soffer, a I member of the NRC Staff who had been a member of the task force, explained:

The major recommendations of this task force were to establish requirements for site approval that were independent of plant design, to try to take into

• consideration the risk of accidents beyond the design basis by establishing population and density distribution criteria and that selected sites should be among the best available in the region, that siting requirements should be stringent enough in the view of the Siting Policy Task Force to reduce residual risk but not so stringent as to eliminate siting from large regions of the country.

Transcript of ACRS Meeting at 45 (January 7, 1992). Mr. Soffer then explained that due to the "low frequency of core damage that is associated with the plants themselves, [b]asically, the safety goal and the ability of the plants themselves is such that they could be sited almost anywhere and meet the safety goal." kl at 47 (emphasis added). In other words, contrary to the Siting Policy Task Force's assumptions, putting the numerical demographic criteria presently in Regulatory Guide 4. 7 (Rev. 1 ) into Part 100 would have no or very little effect on reducing the residual health risk to the population ti&.,_, prompt N*wman & Holtzinger, P. c.

fatalities, genetic effects or excess cancers) over what has been achieved under the current regulatory framework for siting nuclear power plants .

• Within the Commission's Safety Goal decisional framework, substantial increase in protection is based upon consideration of .I§k; namely,
• health effects upon the individual and population ~ supra note 24 regarding health effects objectives). That Is, the measure of whether there has been a substantial increase in protection <.!.&,.., reduction in risk) takes into account .b.Q.:th
• the probability of an accident and the consequences of an accident should one occur. Contrary to the Commission's Safety Goal decisional framework, the
• justification for the proposed changes to the demographic regulations rests, in large part, on consideration of consequences alone independent of their probability of occurrence. In particular, the proposed changes are justified on the basis that
• control of population density out to 30 miles would obviate the need to condemn a large population center (as opposed to less intensively used land} should a very low probability severe accident occur and release cesium or strontium to the environment. SU. Transcript of ACRS Meeting at 70-81 (January 7, 1992).l!f Thus, the basis for the proposed numerical demographic criteria is not so much
• protection of the public health and safety through avoidance of health As ISG Members understand the land condemnation rationale presented by the NRC Staff, the issue Is not so much protection of the population within a condemned population center I from health effects, but avoidance of condemnation (and the attendant property losses) of land that is the site of a large population center. This is because the NUREG-1150 analysis assumes that effective emergency action has been taken. NUREG-1150 at 2-20.

Thus, the residual issue is condemnation of land and loss of economic productivity of that land. By requiring population centers to be located 30 or more miles from a nuclear power plant, any land condemnation would not include a population center, but, at most, land having less intensive use. NtlWmlln & Holtz/ngw, P.C. P.,,.47

consequences {and, hence, defense-in-depth), but avoidance of the need to condemn land on which Is situated a major population center for which the

• emergency actions to protect the population have already been taken.

From the Commission's statements to the effect that plants with a

• smaller exclusion area boundary than 0.4 miles would cause a "very low level of risk" (57 Fed. Reg. at 47,804) and that "nuclear power plants meeting current safety standards could be located at sites significantly more dense than" fut... at
• 47,805) the population density levels proposed, placing 'such stringent demographic limits in the Commission's regulations is not necessary to satisfy the
• Commission's Safety Goals. Moreover, even if consequences alone are relied upon as the basis for justification, such codification would not result in a substantial increase in protection. This conclusion is even more compelling when codification I

carries with it the potential to decrease the benefits accrued from the present flexible application of the demographic guidelines in Regulatory Guide 4. 7 {Rev. 1 ). iii. The adverse Impacts of the proposed revisions greatly exceed their benefits. In addition to being unnecessary to provide adequate protection of the public health and safety or a substantial increase in such protection, the proposed ' revisions to the demographic regulations in Part 100, if adopted, would impose significant adverse impacts (costs), without commensurate benefit. Therefore, in I accordance with the Commission's Safety Goal decisional framework, the proposed revisions, along with draft Regulatory Guide DG-4003 (Proposed Revision Newman & Holt:zlnpr, P.C. Pttg1148

2 to Regulatory Guide 4. 7), should be withdrawn and the current framework left intact .

• The proposed revisions, if adopted, would impose a hierarchy of site characteristics, which elevates demographics over other physical characteristics
• of the site and other safety-related aspects of nuclear power plant siting which may have greater potential for reducing risk. This, in turn, would create the possibility that sites with a better balance overall of favorable safety-related

' characteristics might be eliminated from further consideration on the basis of demographics alone. Thus, adoption of the proposed revisions would upset fundamental, internationally* accepted siting principles directed at selecting sites t based on a careful weighing of site characteristics, including demographic characteristics. Such an outcome would be contrary to the public interest, sound I regulation, and the fundamental safety principles governing the siting of nuclear power plants in the United States and elsewhere in the world. In order to avoid this undesirable outcome, it is necessary that any site safety requirements be flexible enough to take into account demographics, while simultaneously recognizing that demographics alone do not define the risk reduction potential of a site. The present regulations have the necessary inherent flexibility. In 1988, as discussed above, the NRC denied a 1976 petition for rulemaking (PRM-100-2) dealing with siting distances and population densities that would have eliminated flexibility in making site-safety determinations. The Commission found that adoption of PRM-100-2 would have unnecessarily Newman & Holtzinger, P. C. Page4S

restricted NRC' s regulatory siting policies and would not have resulted in a substantial increase in the overall protection of the public's health and safety *

• Another adverse consequence of the proposed regulations is that if the proposed 500 people/square mile population density limit, applied out to 30
• miles, is adopted, otherwise superior sites would be judged to be unacceptable under the proposed rule. The impact of the proposed change can be seen in Figure F9.14 found in NUREG/CR-2239, "Technical Guidance for Siting Criteria
• Development" (November 1982), replicated on the next page. The shaded areas in this figure display locations where the 500 people/square mile limit out to 30
• miles in the proposed rule would not be met as of November 1982. More recent census data would likely show the elimination of larger areas and a greater number of areas. Furthermore, the burden of the proposed rule would not fall evenly across I

the United States. The Mid-Atlantic and New England areas would be most heavily affected. Therefore, the proposed rule would be more restrictive from what is acceptable today. Like PRM-100-2, It could lead to elimination of superior sites. In particular, it would lead to the elimination of sites already approved as superior sites for nuclear power plants.

               , The risk reduction benefit from codification in Part 1 00 of the demographic limits in Regulatory Guide 4. 7 (Rev. 1) would not outweigh the potential for elimination of superior sites.   ~    (the product of the probability of an event times the consequences from the event should it occur) are already very low, as discussed above. This lnherentfy precludes further risk reductions that are substantial. However, the proposed rule discusses consequences as well as~-

Nffmllfn & Holtz/nglK, P.C. P-,,.60

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                                        . ~ ~"'""v~--~-----------------

Even if probability is not taken into account -- an inherent regulatory deficiency -- the proposed rule would not provide any significant reduction in early injury or

• early fatality consequences. Additionally, the proposed rule, if adopted, would result in only a very limited potential decrease in latent fatality consequences.
• Based on the foregoing, it is clear that the loss of necessary flexibility to ensure selection of sites with the most favorable overall characteristics for protection of the public health and safety may compromise fundamental safety
• principles with no or little enhancement of safety. Thus, the adoption of the proposed revisions cannot be justified on grounds of substantial safety
• enhancements. When other significant factors are incorporated into the decision-making process, including the adequacy of present U.S. sites, impact on sites in ISG Member countries and elsewhere, and economic impacts, it is clear that the
• adverse impacts overwhelm any benefit that might result from adoption of the proposed revisions to the demographic regulations in Part 100.

Imposition of the proposed numerical exclusion area and demographic requirements could lead to questions concerning the safety of current nuclear power sites. Although the Federal Register notice on the proposed rule states,

 "[a)n exclusion area of this size [0.4 miles) or larger is fairly common for most power reactors in the U.S.," (67 Fed. Reg. at 47,804), it does not acknowledge that 33% of the present U.S. nuclear power plant sites have an exclusion area smaller than the proposed numerical standard. Similarly, 10% of the present U.S.

nuclear power plant sites exceed the proposed numerical criteria for population t density surrounding the plant site. SECY-92-215 at 6. The situation is similar Newmlln & Holtz/ngl!lr, P.C. l'llf/fl61

elsewhere in the world. In France for example, roughly half of the plants in operation do not meet the proposed exclusion area of 0.4 miles and 5 of the plants

• do not comply with the proposed population density requirement of 500 persons per square mile. In Belgium, Holland, the Rhineland or Luxembourg, the proposed population density and distance requirements, if applied, could exclude all or nearly

' all nuclear power stations. In these areas, the average population density is over 300 persons/sq. km. (750 persons/sq. mi.} -- versus 600 persons/sq. mi. in the I proposed revisions -- and the average distance between major population centers (> than 100,000 persons) is about 50 km. (31 miles). Also, the proposed

• revisions, if imposed, would not only preclude most siting possibilities in these countries, but also raise questions about the location of French power plants on French territory near the relevant borders. In this respect, two of the French sites n9w in operation, which do not meet the proposed population density criterion, are located near the French border with Germany or Luxembourg. A similar situation pertains to Tai~an, where there is also a need to use existing sites for new plants, as well as in Korea.

The comments of Atomic Energy Council of Taiwan to the NRC (February 17, 1993) summarize the likely impact of the proposed revisions on the nuclear power plant siting in other countries: Reactor Siting Criteria (Nonseismic) -- An Exclusion Area Distance of 0.4 miles (640 meters): The distance of the exclusion area boundary for nuclear power plants in Taiwan are 800, 600, 1000, and 350 meters for Chinshan, Kuosheng, Maanshan, and Yenliao sites, respectively. Once the minimum distance of Net/lt/fflan & Holtzlnger, P.C.

I exclusion area is specified explicitly as 640 meter, two sites are already not complied with the revised regulation. It is believed that, although the revision only I applies to the new sites as stated, we are still going to face the challenge from the general public on the related safety issue and spend a great deal of effort In communication and explanation. More than that, due to the limitation of location arrangement, compliance with I this requirement is impossible If adding new units to the existing sites is considered. In other words, the proposed rule change will impose a very big impact, which we think Is not absolutely necessary from the safety point of view, on the development of our nuclear I applications. We would therefore suggest that, instead of requiring a minimum exclusion area distance, NRC place this distance as a recommended value in the Regulatory Guide.

• Reactor Siting Criteria (Nonseismic) - Population Density Criteria:

The population density of 1990, with the unit of person per square mile, within 30 miles of domestic nuclear power plants are as follows: Chinshan Kuosheng Maashan Yenliao 7257 6339 209 6453 It is evident that only Maanshan site can meet this requirement as proposed in the revision of regulation. Again, even though the proposed rule change will not affect the operation of the existing plants as stated, this population density requirement will definitely serve as a strong argument to against the domestic nuclear development. The Korea Electric Power Corporation's comments to the NRC (December 22, 1992) similarly highlighted the difficulties: A. The application of the proposed population density requirement of 600 persons per square mile in Korea will greatly aggravate our ability to acquire suitable sites, which has been a major problem for nuclear power construction due to a public acceptance problem. NIIWffMn & Holtzinger, P. C.

( 1) The average national population density as of 1992 is 1126 persons per square mile, which far exceeds the proposed NRC I requirement. (2) Coastal areas, where the siting of nuclear power plants is most practical and possible is more densely populated than other parts

• of Korea due to the fact that these areas are also suitable for other industrial activities.

(3) Since Korea is an actively industrializing

• country, the projected population density will be even greater in coastal areas .

B. Major Asian countries possessing nuclear power plants such as Korea, Japan, and Taiwan are all densely

• populated, and the proposed regulation will undermine the execution of future projects in these countries as well as other Asian countries.

Population Density Country (persons/sq. mDe) Remarks Korea 1,126 as of '92

**                                   Japan Taiwan 838 1,450 as of '90 as of '90

• C. The numerical demographic criteria will lead to questions concerning the safety of current nuclear power sites which do not meet the proposed population density criteria, not only in the United States but in other countries as well .
• D. There is no current need for codifying demographic criteria because the present Regu,latory Guide 4. 7 works sufficiently for regulatory purposes.

N8Wm1111 & Holtz/npr. P. C. Pllgtt64

Although the applicability of the proposed revisions is explicitly limited to future plants, the fact that large numbers of present U.S. sites would not meet

• limits purportedly relating to site safety and having the force of law raises troubling public acceptance problems about the adequacy of present sites, at the least, and
• quite possibly could provide an arguable basis for petitions and other actions to shut down currently operating reactors.

The proposed codification of numerical exclusion area and

• demographic criteria in Part 100 could adversely impact the siting of future nuclear power plants by unnecessarily limiting the number of potential future sites. Such I unnecessary limitation is especially troubling because of the availability of additional design features to improve, when necessary, plant safety. Within the United States, geographical regions such as the Northeast, which have higher I

relative population densities and less available open land area, may well be precluded from consideration for future nuclear power plant sites if the proposed numerical demographic regulations are adopted. Furthermore, the regulations would preclude siting additional nuclear power plant units at approximately one-third of the presently operating reactor sites in the U.S, even though these sites possess acceptable physical characteristics, important to safe siting. The situation is worse in Western Europe and Asia where size alone limits the availability of acceptable sites. Outside the U.S., particularly in countries with higher population densities and less available land, the public pressure to adopt safety standards similar to the proposed U.S. siting regulations may well make sl~ng additional nuclear plants extremely difficult, if not impossible. ISG Members believe that Newman & Holtzingw, P.C. Page66

I existing sites which are otherwise acceptable should be able to receive new plants with enhanced built-in safety characteristics and margins. I If the rule results in acceptable sites being located far from metropolitan areas, there will be a need for longer transmission lines. In general, longer transmission lines will result in a higher cost for the facility as well as certain disadvantages from power losses which occur over long transmission routes. In some cases, a state or states may find that few, if any, acceptable sites I are available after the new rule is promulgated. If such states must purchase power from utilities located in adjoining states having more favorable sites, there I will be obvious economic impacts on the consumer and the unfavored state. While some favorable economic impacts are also possible, since remote sites may be less expensive to acquire, this possible benefit seems highly theoretical since utilities already strive to obtain the best sites at the lowest possible cost. The foregoing list does not purport to be a comprehensive list of economic impacts. Indeed, through careful review, NRC would doubtless discover many other varieties of economic impacts. 301 The siting rule may operate to favor certain utilities in that their competitive advantage is greatly enhanced. At the same time, the rule could deprive some utilities of any-acceptable sites for nuclear facilities. In some states or regions, the siting rule could operate to render nuclear power uneconomic (or of marginal economic benefit) and thus raise the cost of power in a region or force it to rely on an energy source which has adverse impacts. This may be especially severe in densely populated areas of the country where nuclear power presents a logical answer to high base load demand and where the contamination or pollution from coal or other methods of power generation would be an unacceptable added burden on air standards. Newman & Holtzlnger, P.C. Page 66

B. The Technical Basis for the Proposed Revisions to the Site Safety _Criteria Is Inadequate. lntemaUy Inconsistent and Contusing. The Commission fails to provide an adequate or internally consistent technical basis for including numerical excl'usion area boundary and demographic limits in its site safety regulations in Part 100. The proposed regulations would arbitrarily codify the 0.4 mile exclusion area boundary on the basis of design considerations, but reject the use of dose calculations by fixing the population density limits (500 people per square mile out to 30 miles at the time of initial site approval and 1000 people per square mile 40 years later). The Commission indicates that codification of the exclusion boundary limit in Part 100 would "assure a Y.e..!Y. 1Qw lIDlfil of rjsk to individuals, even for those located close to the plant," 57 Fed. Reg. at 47,804 (emphasis added), and that the population density limits would "meet the Commission's Safety Goals." kl.. at 47,805. However, as discussed in the foregoing sections, codification of these numerical standards in Part 100 would be inconsistent with the Commission's stated policies on regulatory decision-making, in particular, those associated with Safety Goal implementation. Moreover, codification in Part 100 is not necessary to meet the Commission's remote siting goal. Consideration of alternative sites when numerical demographic criteria in Regulatory Guide 4. 7 have been exceeded is sufficient to accomplish this goal, as discussed above. The proposed changes are intended to achieve site isolation through decoupling siting and design, but the rationale for the proposed changes is fundamentally linked to the contributions of design to the reduction of the residual Newman & Holtzlns,t,r, P.C. Page 57

risk from a severe nuclear power plant accident. For example, reference is made to numerous risk studies, such as WASH-1400 and NUREG-1150, which estimate

• risk considering both siting and design characteristics. Further, the justification for an exclusion area distance of 0.4 miles is based on "typical engineered safety
• features." ld... at 47,804. Thus, the proposed revisions rely on past success in design improvements as a basis for achieving site isolation by decoupling siting from design, but refuse to allow credit for future safety improvements to the
• design in siting new plants. Specifically, "the proposed regulation would eliminate the use of a postulated source term, [and] assumptions regarding mitigating
• systems ... " in order to achieve site isolation through decoupling. 57 Fed. Reg .

at 47,804. In sum, the proposed rule is inconsistent because it precludes the use

• of design and risk as a basis for selecting sites, but relies on design and risk as the basis for such preclusion. Additionally, in contravention of the Safety Goal decisional fram~work for imposing safety requirements which are not necessary for adequate protection, it uses the consequences of a very low probability severe accident, as opposed to risk, as the basis for imposing the requirements .
• Further exacerbating the inconsistent technical basis is the Commission's invitation to comment on the size of the exclusion area for plants whose power levels are significantly lower than 3800 MW (thermal). Power level
• is a determinant of the source term, which the Commission would eliminate as a basis for determining exclusion area size. Thus, the proposed rule sends
• conflicting messages on the Importance of source term issues .

Newman & Holtzfngar, P.C. Page 68

As presently constructed, the proposed rule would have site issues affect the determination of the design, but not have design characteristics widen I the choice of sites. Design and risk arguments are used to support certain of the proposed revisions, but, largely based on consequences, revisions would exclude

• the application of design and risk considerations in site selection. Even in this narrower and more inappropriate measure of acceptability, no numerical evaluation of the benefits is offered. When analyses are made of the proposed rule on the I

basis of consequences, the benefits from codification are shown to be either zero or very small. No counterbalancing analysis is provided for the economic and

• health benefits of alternative formulations of regulations which would be consistent with accepted safety principles for siting nuclear power plants.

The inconsistency between the basis for the rule and the requirements

• the rule would impose is made all the more striking when the implications of the proposed rule for the use of existing sites for additional power units or replacement power units are considered. All existing sites are acceptable on the basis of safety considerations. Without grandfathering, however, some existing sites, which would not meet the proposed demographic criteria, would have to be discarded as
• sites for additional or replacement units, even though they are superior sites from an overall safety perspective. Prohibition of gra~dfathering would needlessly result in loss of investment and loss of use of safe sites. Grandfathering, however,
• would likely lead to endless disputes regarding the safety of the existing units.

This example illustrates the flaw inherent in establishing a dual regulatory structure

• in Part 100 where there is no compelling safety rationale for its existence .

Newman & Holtzinger, P.C. Page69

The first of three objectives of the proposed rule is to "[s]tate the criteria for future sites that, based upon experience and importance to risk, have

• been showri as key to protecting public health and safety." 57 Fed. Reg. at 47,803 (emphasis added). However, the proposed rule would impose arbitrary
• prescriptive criteria without any attempt to determine the significance of or necessity for the criteria, such as specified exclusion area distances and specified population densities. In this regard, the Supplementary Information published with
• the proposed changes reflects a misunderstanding of the fundamental purpose of the Safety Goal decisional framework. One of the stated purposes of the Safety
• Goal decisional framework is to provide a better means for testing the adequacy of and need for current and proposed safety requirements . .5..e.e. 51 Fed. Reg. at 28,044. In contravention of this stated purpose, the Safety Goal decisional
• framework is cited as the basis for an unjustifiable ratchet.

Given the progress the nuclear industry has made since 1980 in understanding ~evere accident source terms and the increased application of probabilistic risk assessment techniques to severe accident analysis, the radiological risk to the public is now understood to be considerably different from and less than it was previously thought to be. When coupled with the incorporation of severe accident risk reduction features into new reactor designs, the magnitude of the residual severe accident risk has become small enough that I further reductions in the residual risk from siting on the basis of demographics will be very small. To be judged adequate and credible, particularly in light of the Nnnnan & Holtzinger, P. C. Page60

Commission's rulemaking objective quoted above, the technical basis must reflect such facts. C. The Proposed Revisions to the Seismic Criteria Should Not Be Adopted, Revision Should Await Resoludon of the Controversy on the Use of Deterministic Versus ProbabDistlc Methods In Site Selection. Any Revision Adopted Should Meet the Commission's

• Rulemakjng Objective of Regulatory StabQity, The Commission issued seismic and geologic siting criteria in 1973 in the form of Appendix A to Part 100 (38 Fed. Reg. 31,279 (November 13,
• 1973)). At the time of their issuance, the seismic criteria reflected state-of-the-art understandings in the conduct of seismic and geologic investigations and were developed with the cooperation of the U.S. Geological Survey and the National Oceanic and Atmospheric Administration. ld.i. at 31,280.

The primary reasons given in the Federal Register notice for the proposed changes to the seismic and geologic. criteria are (1) to benefit from experience gained in the application of the procedures and methods set forth in the present regulatipns; (2) to incorporate the rapid advances in the earth sciences and earthquake engineering that have been made since the criteria were first published in 1972; and (3) to reduce the difficulty encountered by nuclear power plant applicants and the NRC Staff in exercising needed judgment in applying the criteria and using evolving methods of analyses within the context of the licensing process, thereby leading to a more stable and predictable licensing process than in the past. ~ 57 Fed. Reg. at 47,803. The ISG does not believe that the proposed revisions to the present criteria in Part 100 will accomplish these objectives. To the contrary, the proposed revisions could lead to greater instability Newman & Holtzlnger, P.C.

and less predictability in the licensing process than under the present regulatory regime. In view of the potential attractiveness of other alternatives and the lack

• of a resolution methodology should deterministic and probabilistic methods give divergent results, the prudent course is for the Commission to withdraw the
• proposed revisions .

1. The Commission should resolve the controversy between use of deterministic versus probabilistic techniques before proceeding to rulemaking .

• The Supplementary Information published with the proposed changes makes clear that the present, deterministic approach "has worked reasonably well
• for the past two decades, in the sense that SSEs [Safe Shutdown Earthquakes] for plants sited with this approach are judged to be suitably conservative, [even though] the approach has not explicitly recognized uncertainty in the geoscience
• parameter." ld.z. at 47,807. Because the NRC wants to require use of probabilistic methods, which allow treatment of this uncertainty, the revisions are being proposed. The J?roposal to use a dual scheme by adding probabilistic methodology to the existing deterministic methodology for seismic design criteria will unnecessarily complicate and destabilize decisions concerning selection of design
• basis ground motion and, consequently, site selection. Controversy over the appropriate probabilistic methodology for seismic analysis has led to the development of two distinctly different approaches in the U.S. As the NRC acknowledges, "[b]ecause so little is known about earthquake phenomena (especially in the United States) . . . [e]xperts often delineate very different
• estimates of the largest earthquakes to be considered and different ground-motion Newm11n & Holtzinger, P. C. Pas,e 62

models." ~ Application to seismic analyses of a probabilistic methodology that has not yet been fully evaluated and tested through actual use in the licensing

• process results in even greater controversy. In particular, the bottom-line results from probabilistic seismic hazard analyses tend to be dominated by the extremes
• rather than the central tendencies of the distributions of knowledge and expert opinions. llL. This controversy among seismic experts, coupled with the divergent views within the NRC Staff as to the role probabilistic seismic hazard analysis
• should play in the licensing arena, will undoubtedly have a destabilizing, rather than a stabilizing, effect on the siting process.
• Given the controversy surrounding the appropriateness of using such analyses in the licensing process, the lack of experience on the part of both the NRC Staff and nuclear power plant applicants in such use in the licensing process,
• and the availability to the Commission of other, more suitable means to gain experience with the application of probabilistic methodology to seismic analyses (such as the use of pilot programs or issuance of a policy statement), there is no valid reason for proceeding with these revisions.

At a minimum, if the Commission insists upon codifying a

• methodology requiring that both deterministic and probabilistic methods be used in nuclear power plant siting, it should identify a resolution methodology to be applied whenever the two kinds of studies produce divergent results. Without
• such a methodology, the revisions requiring use of both deterministic and probabilistic studies will destabilize the licensing process and introduce greater unpredictability into it.

Newman & Holtzlnger, P.C. Page 63

2. The proposed requirements are unlikely to meet the stated objectives of greater predictabHlty and stability .

• There is not consensus within the NRC Staff as to whether or how probabilistic analyses should be used in ~aking siting decisions. 47 Fed. Reg. at 47,812. Moreover, the application in the licensing process of such analyses to
• nuclear power plant siting has not yet been tested. Lack of consensus and lack of a body of licensing practice are destabilizing forces in the licensing process.
• Under these circumstances, it cannot be expected that adoption of the proposed revisions to the seismic criteria will lead to greater predictability and stability.

Imposition at this time of seismic criteria in the form proposed is I especially inappropriate. In the seismic-hazard area, the Commission has stated that the bottom-line results from probabilistic analyses tend to be dominated by the extremes rather than the central tendencies of the distributions of knowledge and I expert opinions. Also, "[b]ecause so little is known about earthquake phenomena (especially in the eastern United States), . . . [e]xperts often delineate very different estimates of the largest earthquakes to be considered and different ground-motion models." kt:. at 47,807. Additionally, there are divergent views within the NRC Staff as to the role probabilistic seismic hazard analysis should play In the licensing arena. In particular, "views range from an advocacy of a predominantly probabilistic analysis to the probabilistic/deterministic analysis [being] proposed [in the revisions] to a predominantly deterministic approach as used currently." ld... at 47,812. Newman & Holtz/nglJI',. P.C. P.,,,.64

' Due to the divergence of views among the NRC Staff, the Commission is requesting comments on specific questions. The Commission has I . asked whether deterministic and probabilistic evaluations should be combined or weighted, whether the procedure specified In one of the draft Regulatory Guides I to determine controlling earthquakes from the probabilistic analysis is adequate, whether median values of the seismic hazard analysis should be used to the exclusion of other statistical measures, whether the exceedance criterion for the I Safe Shutdown Earthquake Ground Motion is properly specified, and how many earthquakes should be generated to cover the frequency bands of concern for

• nuclear power plants . .5.@i.d.i.at 47,812-13. As discussed above, the controversy should be resolved before adopting new requirements, due to the fundamental problems remaining in the criteria as proposed. For example, further effort should I

be expended in developing the earthquake database, improved techniques for weighting the seismic zone area, and improved expressions for the attenuation equation. At a .minimum, no rule should be proposed which does not resolve all issues which could lead to instability in_ the licensing process. Even though the present criteria are not perfect, their operation is understood by both applicants and the NRC. The proposed revisions are fraught with regulatory uncertainty of a fundamental nature and should not be finalized until further evaluation is completed. Newnum & Holtzlnger, P. C.

I

3. Any revision to the Commission's seismic regulations should provide regulatory stability.

I As discussed above, there is not agreement among the NRC Staff as to how probabilistic analyses should be used in siting nuclear power plants. This Is because neither the NRC Staff nor the nuclear Industry has developed a body of I practice, in the licensing context. As the Federal Register notice implicitly acknowledges, the nuclear regulatory experience is limited to those who I participated in either the NRC-Lawrence Livermore National Laboratory or the Electric Power Research Institute seismic hazard research projects over the last decade. ~ llt. Additionally, under the sponsorship of the Nuclear Management I and Resources Council (NUMARC) an alternative has been developed. Therefore, it is both appropriate and prudent' for the Commission to proceed cautiously with codification of requirements in the absence of licensing I experience with implementation of the requirements. In similar instances in the past, the Commission has employed a trial approach in order to develop the necessary body of practice within the regulatory context. Such an approach is especially appropriate where, as is the case here, the proposed revisions have limitations and there is significant controversy among experts as to whether and I how the probabilistic criteria should be applied. For example, in implementation of the Policy Statement on Safety Goals, the Commission proceeded cautiously I and, in the area of severe accident regulation, it is only now embarking on a rulemaking based on years of safety research. 57 Fed. Reg. at 44,513 (September 27, 1992). Also, NUMARC's proposal appears to merit evaluation by the I NflWJTla/1 & Holtzlnger, P. C. Page fSfS I

I Commission before the Commission makes a final decision on the structure and content of revised siting criteria. I The Federal Register notice for the proposed revision to the seismic criteria asserts without elaboration that "the NRC believes that this approach is the I ~ way to accomplish the objective, ... and arrive, through analysis, at a site-specific ground motion that appropriately captures what is known about the seismic regime"; and that the approach "should lead to a more stable and I predictable licensing process than in the past." 57 Fed. Reg. at 47,807 (emphasis added). Two aspects of this assertion need to be addressed. First, rulemaking I does not have to be used to achieve the desired result. Second, the approach taken in the proposed rule has not been shown to be the~ approach. As to the first point, the Commission does not have to modify the

• existing Part 100 seismic and geologic criteria in Appendix A to Part 100 in order to obtain the submittal of probabilistic analyses in addition to deterministic analyses as part of the assessment of the seismic and geologic properties of a site.

Applicants who wish to conduct probabilistic analyses and submit them to the NRC in conjunction with their deterministic analyses are certainly not prohibited from doing so and the NRC Staff can indicate to applicants its interest in such analyses. If the Commission wishes to require such probabilistic analyses, it I could issue a policy statement requiring the submittal of probabilistic analyses. This would be similar to the approach taken in the Commission's Severe Accident Policy Statement (50 Fed. Reg. 32,138), which requires that a Probabilistic Risk Newnum & Holtzinger, P.C. Pas,e67

Assessment (PRA) be completed for all new plants and included with the application. This PRA must be used to expose the severe accident vulnerabilities

• initiated by both internal and external events that are associated with a plant of new design. For the Severe Accident Policy Statement, the NRC Staff must
• complete a review of a PRA as part of its licensing review of a design for a new nuclear power plant. Although a policy statement does not establish a requirement as a legal matter, it has the practical effect of a requirement. In conjunction with issuing such a Policy Statement on probabilistic seismic analyses, the NRC Staff might issue guidance as to how the Policy Statement would be implemented for
• both the Staff itself and applicants .

As to the second point, the Commission has not provided any justification that the proposed ~evisions to the seismic criteria are the .b.§fil;. To the

• contrary, the Supplementary Information for the proposed revisions demonstrates that there is controversy concerning their formulation. Additionally, in its comments on ~e proposed revisions to Part 100, NU MARC has proposed an alternative which appears to warrant evaluation. At a minimum, the Commission should evaluate the NUMARC proposal (and any other attractive alternatives) and resolve the present controversy before finalizing any changes. Such an evaluation would also permit the Commission to make further progress in such areas as development of an earthquake database, Improved techniques for weighting the
• seismic zone area, and improved expressions for the attenuation equation. In view of the infirmities in the rule as a whole, the ISG urges the Commission to withdraw the rule in its entirety until these issues are adequately addressed.

Newman & HoltzbJf/td, P. C. Page68

D. The Environmental Assessment Prepared lo Conlunction with the Proposed Revisions Is Inadequate as a Matter of Law to Support a Finding of No Significant Environmental Impact .

• The Commission's regulations In 10 CFR Part 51 specify the procedures necessary for compliance with NEPA. At a minimum, Part 51 requires
• that an Environmental Assessment (EA) be prepared examining the environmental impacts of the proposed action and reasonable alternatives to the proposed action.

The purpose of an EA is to determine whether a comprehensive Environmental Impact Statement (EIS) must be prepared. Section 51.30 requires that an EA identify the proposed action and include:

• ( 1) A brief discussion of:

(i) (ii) The need for the proposed action: Alternatives as required by section 102(2)(E) of NEPA; (iii) The environmental impacts of the proposed action and alternatives as appropriate; and

• (2) A list of agencies and persons consulted, and identification of sources used.

The EA prepared in conjunction with revision of the Commission's siting criteria fails to meet the Commission's NEPA requirements in Part 51 and is otherwise not in accordance with law. The proposed revisions to the 10 CFR Part 100 siting regulations will result in substantive and significant changes to the Commission's regulatory framework for siting new nuclear power plants, if adopted. Specifically, the demographic criteria as formulated In the proposed rule do not ensure that the sites chosen will be among the best reasonably to be found. Given the importance of the U.S. siting regulations and guidance to the formulation of international site NtlWmlUI & Holtzinger, P. C. Pilg* 69

safety standards, adoption would force reconsideration of present international safety standards and raise questions about the adequacy of present siting I r: practices. Adoption could have the practical effect of making it difficult, if not impossible, to site nuclear power plants in any portion of the United States which

• did not meet the numerical demographic criteria specified in the regulation, as well as elsewhere in the world. Adoption also has the potential to create confusion among members of the public as to the adequacy of existing nuclear power plant I

sites and the adequacy of existing emergency planning requirements. The EA fails to address the issue of multi-unit sites and the economic impacts of the regulation

• on siting decisions. These are the kinds of environmental impacts which NEPA requires be addressed and taken into account as part of the process to decide whether to make the proposed revisions final.

I The inadequacy of the current EA is corroborated by the fact that in 1980, the Commission determined that it would prepare an Environmental Impact Statement (EIS) in connection with revision of* its siting criteria to incorporate numerical demographic criteria into a revised 10 CFR Part 1OO.ll' In support of On July 29, 1980, the NRC issued an Advance Notice of Proposed Rulemaking (ANPR) (45

• Fed. Reg. 50,350), in which the Commission announced Its intention to revise the reactor siting criteria and requested comments on seven of the nine recommendations of the Siting Policy Task Force, as well as certain alternative approaches. In conjunction with the rulemaking effort, the Commission also Issued a Notice of Intent (NOi) to prepare an Environmental Impact Statement (EIS). 45 Fed. Reg. 79,280 (December 2, 1980). The NOi, among other things, Identified the technical approach to detailed analyses that would I be followed in developing the bases for any proposed revisions. ~ kL. at 79,822-23. In December 1981, the NRC published the Scoping Summary Report for the EIS (NUREG-0833). The report addressed comments received on both the ANPR and the NOi and provided further discussion of the efforts which would be undertaken to develop an adequate technical basis for any revisions. The report recognized that the siting rulemaking must take into account premises concerning reactor design and emergency planning. fum I (continued ... )

Newman & Holtzinger, P.C. Pas,e70

the development of the EIS, the NRC identified major studies to be undertaken to understand the impacts. Significant changes have occurred since issuance of the

• Notice of Intent and Scoping Summary Report in Commission policy, practice and capability to determine how It will proceed in establishing requirements, such as
• the proposed changes to the siting regulations, when such changes are -not necessary to ensure adequate protection of the public health and safety. However, the kinds of studies'identified remain as valid in 1993 as they were in 1980 for
• assessing the impacts of the proposed regulation and serving as the basis for an EA. For example, the following studies were identified in the 1980 Notice of
• Intent ~ 45 Fed. Reg. at 79,822-23):

                   <1 > Radiological Consequences of Accidents: Proposed criteria will be compared with realistic alternatives on the basis of impacts on public health and safety. For demographic criteria this means that variation in doses

• to the maximally exposed individual and the population from a full range of accident releases must be examined for alternative ways of specifying constraints on population density and distribution. Existing sites and a t:,ypothetical site will be evaluated. Consequences considered will include early fatalities, injuries, latent fatalities, and property damage. Both individual and societal risk will be evaluated but may differ in relative importance for establishing different criteria.

(2) FeaslbHity of Protective Actions: The topics under consideration for rulemaking with respect to demographic criteria and external hazards will be examined to determine whether the capability to take protective action in the vicinity of a site under

• W( **. continued)

NUREG-0833 at 13. It also restated that *a systematic evaluation of accident consequences for a full range of reactor accidents would be a fundamental part of the technical basis for the siting rulemaking . . . .

• jg_,_ at 20. See aiso ~ at 14, regarding consideration of a full range of accidents in establishing siting criteria.

Newman & Holtzinger, P.C. Pap71

accident conditions might be impaired or enhanced by various choices of alternative criteria .

• (3) Definition of Region: Alternative schemes of regionalization will be examined to determine a proper basis for establishing regional criteria. Socioeconomic and physiographlc units will be examined to establish potential regional breakdowns. Effects of uniformity of
• population distribution, water resource restrictions and any other appropr~ate regional concerns will be considered when deciding on the proper regionallzation scheme.
• (4) Site AvaDabUlty: Consistent with the intent of the NRC FY-80 Authorization Actw, the new In June 1980, the U.S. Congress passed the NRC Authorization Act for Fiscal Year 1980 (FY-80), Pub. L. No. 96-295, 94 Stat. 780 (1980). Section 108 of the Act provided in
• pertinent part that:

(a) . . . mhe Nuclear Regulatory Commission is authorized and directed ... to develop and promulgate regulations establishing demographic requirements for the siting of utilization facilities . . . *

• (c) The regulations . . . shall specify demographic criteria, including maximum density and population distribution for zones surrounding the facility without regard to any design, engineering, or other differences among such facilities.

The Conference Report (H.R. Cont. Rep. No. 96-1070, 96th Cong., 2nd Sess. 24 (1980)), linked the legislation to the Siting Policy Task Force recommendation to strengthen siting as a factor" in defense-in-depth, but without eliminating further siting of nuclear reactors in any region of the United States. ,S,ftft H.R. Cont. Rep. No. 96-1070 at 25-26. Like the Siting Policy Task Force Report upon which It relied, the FY-80 NRC Authorization Act reflected understandings of the time that codification of remote siting requirements in the Commission's regulations would contribute significantly to reducing the residual risk from a severe nuclear power plant accident and, hence would contribute to increased defense-in-

• depth.

Section 108 of the Fiscal Year 1980 Authorization Act has long since expired, as evidenced by its not having been codified in the U.S. Code and Its having no language indicating permanency. In Massachusetts y. NRC, the Court referred to the FY 80 NRC Authorization Act as *an expired fiscal appropriations law, [that) was not in effect when

• CLl-90-2 was decided and therefore did not limit the licensing discretion otherwise conferred on the Commissiqn by Congress.* Massachusetts v. NRC, 924 F.2d 311, 324 (D.C. Cir. 1991 ). However, the assessment identified in the 1980 Notice of Intent to

       , ensure that new demographic regulations do not preclude further siting of nuclear power plants in any region of the United States is still needed, the more so because changes to the demographic regulations In Part 100 are not needed to ensure adequate protection of I         the public health and safety.

Newm,,n & Holtzlnger, P.C. ht,e72

I demographic criteria should not preclude further siting of nuclear power plants in any region of the United States. An assessment will be made for each region I that identifies the variation In availability of sites for nuclear power plants as a function of the structure of the criteria and the variation in numerical values as well as realistic constraints on siting such as water ayailabjlitv and violation of safety criteria. The benefits I of regionally based criteria versus nationwide criteria will be examined. Basic Information will be developed from existing siting studies which, taken together, cover large portions of the country. (Emphasis added.) I (4) Socioeconomic Impacts: The socioeconomic impacts of varying degrees or remoteness will be investigated. Economic impact of increased transmission distances, impacts on land use and other factors will be addressed along with sociological penalties and inequities in distribution of cost and I benefits of such siting. (5) Severity of External Hazards: A literature review will be performed to establish the potential level of hazard associated with the external hazards listed in the

• [ANPR] and any other appropriate topics. Staff practice for dealing with these hazards will be assessed.

Available models for characterizing the effect of a hazardous external event will be evaluated. The feasibility of establishing a meaningful protective distance will be examined. The availability of sites associated with the demographic criteria proposed by the staff will be reexamined to determine whether the standoff criteria will significantly alter site availability .

*                  (6) Engineering Alternatives to Standoff Distances: The feasibility of design performance requirements as opposed to specific standoff distances will be evaluated.

(7) Precludlna Siting of Nuclear Reactors In Any Region

• of the United States: Energy generation from any source has its associated risk and risks from some energy sources may be greater than that of the nuclear option. Therefore, it has been suggested that the siting criteria should not be so stringent as to preclude the use of nuclear power from any region of the United States.

Newman & Ho/tzlnger, P.C.

I The implications of not precluding nuclear power from any region of the United States will be examined. I (8) Effect of Groundwater Interdiction Criteria on Site Availability: The effect of site availability of alternative siting criteria that assure the capability for groundwater interdiction would be examined. I (9) Use of Existing Sites: The existing sites would be examined for various levels of criteria to determine which sites were acceptable under each proposal. The feasibility of adding additional units to each of these - sites would then be examined and an estimate made by I region of remaining siting capacity. ,Using the characteristics of the selected site, an estimate would be prepared of the availability of multi-unit sites as a modification of the availability information for the various demographic criteria and standoff distances .

• ( 1O) Use of Unusual or Unproven Engineering Design to Compensate for Site Deficiencies: An estimate would be made of the effect on site availability of instituting such a requirement, particularly where large areas might
• have a common deficiency which might preclude siting from a large region.

The economic aspects of reactor siting have long been a fundamental part of the NRG's NEPA review. In numerous cases the Commission's tribunals have considered the economic aspects, among other factors, of alternative facility sites. 331 Since the NRC has consistently interpreted NEPA as requiring it to I assess cost-benefit matters, including economic factors, in individual adjudications concerning construction permits and operating licenses, the NRC is plainly under a duty to inquire into the cost-benefit aspects of its proposed siting rule. The need

• for a comprehensive and accurate inquiry into economic impacts cannot be

*          ~  Union of Concerned Scientists y. AEC. 499 F.2d 1069, 1084-85 (D.C. Cir. 19741 .

Newman & Holtzins,er, P.C. Page74

overstated. Since acceptable sites for nuclear power plants are difficult to find and costly to acquire, a siting rule is certain to have significant Impacts on the costs

• of constructing a nuclear facility! An Irony concerning the cost of nuclear facilities is that costs have risen sharply in response to tightened NRC engineering safety
• requirements. Yet through this proposed rulemaklng, the NRC would ignore many of the engineered safety features which have been required for plants at considerable cost. Some of the more obvious economic impacts of such a siting
• rule are discussed above in Section 11.A.2.c.iv.

An agency's Finding of No Significant Environmental Impact, and

• hence the adequacy of the EA which provides the basis for the finding, is judged against a standard of reasonableness. ~ Natural Resources Defense Council v.

Duvall. 777 F.Supp. 1533, 1537 (E.D. Cal. 1991 ). In determining adequacy,

• courts have considered, among other things, the following questions:

Has the agency accurately identified the relevant environmental concern 7 Once the agency has identified the problem, has it taken a "hard look" at the problem in preparing the EA 7 If a finding of no significant impact is made, will the agency be able to make a convincing case for its finding 7 I

 ~    Sierra Club v. DOT. 753 F.2d 120, 127 (D.C. Cir. 1985) (citations omitted).

I The draft EA prepared in connection with the proposed revisions to the siting regulations would not be found adequate under this test. In its 1980 ANPR, the Commission sought public comment on substantially the same revisions Newmen & Holtzlnger, P. C. Page 76 I

I to the Commission's demographic regulations in Part 100. The Commission received a number of comments on the ANPR. Commenters emphasized that I current siting practices were effective in achieving isolation, with the trend being toward the siting of nuclear power plants away from highly populated areas. I Commenters expressed concern about the inadequacy of the technical basis for the changes under consideration and requested that the Commission develop safety goals, quantify residual risk and establish an overall risk criterion before I proceeding. Several commenters thought that an EIS should be prepared, which would consider among other things, the disadvantages of remote siting and the

• elimination of new nuclear power plants from certain regions of the country should the contemplated revisions be adopted. The potential impact on worldwide siting was also called to the Commission's attention.

I The Commission's Advisory Committee on Reactor Safeguards (ACRS) also commented on the issues presented in 1980 when the Commission issued the ANPf:i. While the ACRS agreed that siting, as a factor in the defense-in-depth philosophy should be strengthened, the ACRS stated: mhe ACRS believes that any minimum requirements for parameters such as the exclusion zone radius, I surrounding population density, or distance from population centers should be established, if possible, within the framework of an overall Nuclear Regulatory Commission safety philosophy for future reactors.

• Such a philosophy should be based on preestablished Commission objectives for acceptable risk to both individuals and society. This will, of necessity, include consideration of matters such as the potential effects of a broad spectrum of reactor accidents, the identification of ALARA (As Low As Reasonably Achievable) criterion lwNman & Holtzinger, P.C. Page 76

I for the reduction of risk from accidents, and a general statement of policy concerning the objectives to be sought in reactor design with regard to the prevention I and mitigation of accidents. The establishment of demographic-related site criteria will inevitably require a considerable amount of judgment. However, the choice will be less arbitrary if I made within the framework of an overall NRC safety policy. 45 Fed. Reg. at 50,352. An Errata Sheet to the 1981_ Scoping Summary Report, which I addressed comments received on both the ANPR and Notice of Intent, indicated that the Commission would re-examine its decision to prepare an EIS when it resumed the rulemaking to revise the siting regulations. ISG member~ do not I believe issuance of the draft EA without explanation of why the Commission , apparently changed Its mind about preparing an EIS is consistent with case law. An agency's decision not to proceed with an EIS is unreasonable if the agency I

  "fails to supply a convincing statement of reasons why potential effects are insignificant."     Seattle Community Council Federation v. FAA, 961  F.2d 829, 832 (9th Cir. 1992) _(citation omitted). An agency's decision is also unreasonable if substantial questions are raised regarding "whether the proposed action .ma.y have a significant impact upon the human environment."          kl.. ~ iil.§Q Blue Ocean I

Preservation Society y. Watkins, 767 F. Supp. 1518, 1526 (D. Haw. 1991 ). The Commission's seeming failure, in preparing the EA for the 1992 proposed I revisions, to take into account the comments generated in response to the 1980 ANPR and Notice of Intent cannot be considered reasonable, given their continuing Newman & Holtzlnger, P.C. Pap77

relevance to the changes proposed. Ignoring those earlier comments is not a "convincing statement."

• In sum, the Environmental Assessment prepared in conjunction with the proposed revisions to the Commission's siting regulations is inadequate as a
• matter of law to support a finding of no significant environmental impact .

E. Finalization of the Proposed Revisions to the Siting Regulations Would Not Be in Accord With Sound Agency Declslonmaldng, I As the Supreme Court held in Vermont Yankee Nuclear Power Corp,

v. Natural Resources Defense Council. 435 U.S. 519 (1978), administrative agencies have reasonable latitude in the rulemaking process, and the courts will

• not impose their own notion of what procedures are best or most likely to further the public good. Still standing, however, is the requirement that an agency avoid conduct which is arbitrary or capricious, and that it provide a reasonable statement I

of the basis and purpose of its rules. Moreover, an agency must obey its own regulations. On many occasions the courts have struck down agency rules which were adopted without adequate agency review of major matters related to the rulemaking. The NRC's present course is perilously close to the type of conduct which courts frequently strike down. I The siting rule, if promulgated, may well be vulnerable to legal challenge on the grounds that the Commission's analyses and decisionmaking have I omitted consideration of critical information in the formulation of the proposed revisions, even though such information was available to the Commission as a result of the earlier issuance of the ANPR and Notice of Intent. Furthermore, the Newmv, & Holtz/nger, P.C. I

proposed rule lacks an adequate technical basis, and the Environmental Assessment required by NEPA is patently inadequate. Additionally, the analytic I process prescribed by the Commission's Safety Goal Policy Statement and implementing guidance has not been followed in formulating the technical basis. I Given the importance of the siting rule, it is critical that the Commission closely observe prudent administrative practices. Too much is at stake here for the Commission to proceed further when important aspects of the siting rule are not yet developed. Following publication in the Federal Register of the Advance Notice

• of Proposed Rulemaking in July 1980, the NRC supplemented the notice in December 1980 with its Notice of Intent to Prepare an Environmental Impact Statement. The following year, in December of 1981, the Commission "deferred" I its rulemaking process concerning siting criteria to await development of Safety Goals and improved research on accident source terms. Finally, on October 20, 1992, the NRC_issued a proposed rule concerning reactor siting criteria.

The NRC's decision to publish the ANPR began the rulemaking process regarding siting criteria. Indeed, the NRC's October 20, 1992, Federal

• Register notice states that the rulemaking process began over twelve year~ ago.

57 Fed. Reg. 47,802. Although the 1986 Regulatory Agenda and 1988 denial of PRM-100-2 indicate that the NRC's Executive Director for Operations had I concluded the 1980 rulemaking should be terminated, the Commission, by its own admission in the 1992 Federal Register notice, clearly regards the rulemaking as ongoing since 1980. Consequently, the NRC, in formulating the proposed Page79

I revisions to the demographic regulations, should have considered the comments submitted in response to the ANPR issued on July 29, 1980, as well as the

• December 2, 1980 Notice of Intent to prepare an EIS and the responses thereto by the NRC Staff in the December 1981 Scoping Suml)larv Report.

I Under Section 553 of the Administrative Procedure Act (5 U.S.C.

  § 552 .m; .§ruh (1988)), an agency must "incorporat~ in the rules adopted a precise general statement of their basis and purpose."        5 U.S.C. § 653(c) (1988).      A failure to address substantive comments or concerns raised during the rulemaking renders tHat process invalid. Marsh v. Oregon Resources Council, 490 U.S. 360, I 378 (1989); Bethesda Hosp. y. Heckler. 609 F. Supp. 1360, 1371 (S.D. Ohio 1985). If, upon review, a court cannot discern that an agency made a reasoned decision after consideration of relevant factors, the agency action is art;>itrary and

• capricious. kl.. As discussed above, the Commission's failure thus far to address relevant comments in proceeding from one stage of the rulemaking to the next, in the process of ~evising its siting regulations, has the potential to render any rule which may be adopted vulnerable to invalidation.

Also, an agency is also bound by its own regulations. Robert E.

• Pereckter of Rhode Island. Inc, v. Goldschmidt. 506 F. Supp. 1059, 1063 (D.R.I.

1980). The NRC's NEPA regulations in Part 61, described above, require the NRC to explain the basis for its actions. To the extent that the Commission has ignored

• its earlier ANPR and NOi actions, the NRC is out of compliance with its own regulations. Similarly, the technical basis for the proposed revisions to the demographic regulations is inconsistent with the Commission's Safety Goal Newman & Holtzlnger, P.C. Pat,eBO

I decisional framework. The problems with the proposed revisions to the siting regulations and the rulemaking process leading up to their proposal are so great

• that withdrawal of the proposed revisions and termination of the proceeding seems the most prudent course .

Ill. CONCLUSIONS For the reasons set forth above, ISG Members believe that:

• 0 The existing demographic regulations in 10 CFR Part 100 have worked well; 0

The proposed revisions to the demographic regulations

• 0 are unnecessary; Contrary to the Supplementary Information published with the proposed rule, the proposed revisions to the demographic regulations do not codify present siting
• practice, but change practice in a fundamental way such that there no longer can be assurance that a site proposed for a nuclear power plant will be among the best reasonably to be found; 0

The proposed revisions to the demographic regulations are unduly restrictive and without commensurate bene~t; and 0 Adoption of the proposed revisions to the demographic regulations will have adverse consequences on the

• internationally accepted consensus standards of the IAEA on the siting of nuclear power plants and on national standards in ISG Member countries.

Consequently, ISG Members urge the Commission to withdraw the proposed

• revisions, along with draft Regulatory Guide DG-4003 (Proposed Revision 2 to Regulatory Guide 4. 7), to the demographic regulations in 10 CFR Part 100.

Newm,,n & Ho/tzinger, P.C. PlkgeB1

Likewise, ISG Members request the Commission to withdraw the proposed changes to the seismic criteria in Part 100. Withdrawal would allow I resolution of present controversies concerning the proposed changes and evaluation of alternatives. Withdrawal and evaluation of alternatives would provide I a better basis for the development of international consensus standards reflecting the principles embodied in the NRC regulations . I I I Newman & Holtzlnger, P.C. /',ige82

I ApPENDIX

• In addition to soliciting comments on all aspects of this rulemaking, the Commission, in Section XI of the Federal Regjster notice requested comments on a number of questions. The International Siting Group's responses to these
• questions are found in this Appendix .

A. REACTOR SITING CRITERIA (NONSEISMIC)

1. Should the Commission grandfather existing reactor sites having an

• excluslon area distance less than 0.4 mHes (640 meters) for the possible placement of additional units, H those sites are found suitable from safety consideration?

ISG Response: This question presupposes that the Commission

• will revise the existing reactor siting regulations to include a numerical size requirement, in terms of distance, for the exclusion area. The International Siting Group (ISG) does not believe that it is
• either necessary or desirable to change the existing siting regulations.

As discussed In Section ll(A)(1) of the ISG Comments on the proposed revisions to the siting regulations, it is essential for siting standards to be sufficiently flexible to "ensure that all site-related characteristics have been taken into account" during the selection of

• the preferred candidate sites. See IAEA Safety Guide No. 50-SG-S9, Site Survey for Nuclear Power Plants ( 1984) at 10. That guide
• identifies fourteen (14) safety-related site characteristics to be evaluated during the site selection process, of which population
• - A1 -

distribution Is but one.!' The guide recognizes the difficulty in comparing sites based on population and suggests that "[i]t may be appropriate to compare all other site characteristics, and then to evaluate the sites independently from the point of population distribution. n ld... at 32. Contrary to what is essential in the selection of preferred sites,- the proposed revisions, if adopted, would impose a hierarchy of site characteristics, which elevates demographics over other physical characteristiqs of the site and other safety-related I aspects of nuclear power plant siting which may have greater I potentia* for reducing risk. This, in turn, creates the possibility that sites with a better balance overall of favorable safety-related characteristics may be eliminated from further consideration on the

• basis of demographics alone. Such an outcome would be contrary to the public interest and sound regulation and the fundamental safety

J The other thirteen are:

            -  Surface faulting
            -- Seismicity
            -- Suitability of subsurface material I            -  Vulcanlsm
            -  Flooding
            -- Extreme meteorological phenomena
            -  Man-induced events
            -  Dispersion in air
            -  Dispersion in water I            -  Emergency Planning '
            -  Land use
            -  Availability of cooling water
            -- Other site characteristics as appropriate, such as avalanche, landslide, surface collapse.

I M.. at 10-13.

                                           * -A2 -

I

principles governing the siting of nuclear power plants everywhere in

• the world .

Regulations which grandfather are always problematic. They

• establish a dual system of seemingly conflicting standards and are confusing to the public. Grandfathering conveys to the public the message that what is grandfathered is less safe than what is not I

grandfathered. Once this occurs, it is very difficult to convince the public otherwise. In the case of the proposed revisions to the

• demographic requirements in the Commission's siting regulations, there is no need to introduce grandfathering clauses into the Commission's demographic requirements because there is no need to I

change the requirements at all. As discussed in Section ll{A)(2)(a) of the ISG Comments on the proposed revisions to the siting regulations, the existing siting regulations have achieved site isolation. As discussed in Section ll(A)(2)(b), adoption of the proposed revisions is not needed to ensure "decoupling" of nuclear

• power plant siting and design. As discussed in Section ll(A)(2)(c),

adoption of the proposed revisions to the demographic regulations in

• Part 100 will not provide a substantial increase in protection nor contribute to increased defense-in-depth. Further, the adverse impacts of the proposed revisions greatly exceed their benefits. And,
• - A3 -

I as discussed in Section 11(8), the technical basis for the proposed

• revisions to the siting criteria is inadequate, internally inconsistent and confusing.

I

2. Should the exclusion area distance be smaller than 0.4 mile (640 meters) for plants having reactor power levels slgnHlcantly less than 3800 Megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example, 0.25 miles or 400 meters)?

I ISG Response: See ISG Response to Question 1 above. The exclusion area should be allowed to vary according to power level, as

• is the case with the current regulations. In this regard, power level is a determinant of the source term which is used in setting exclusion area size .

The question illustrates a basic inconsistency In the technical basis

• - fo~ the proposed rule. The proposed-rule, if adopted, would eliminate source term as a basis for determining exclusion area size.

Question 2, however, implicitly suggests that consideration of the

• source term is an appropriate way to determine exclusion area size .

                               -A4 -

3. The Commission proposes to codify the population density guidelines In Regulatory Guide 4. 7 which states that the population density

• should not exceed 500 people per square mDe out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter. Comments are speclflcally requested on questions 3(a), 3(b), and 3(c) given below.
• (a) Should numerical values of population density appear In the regulation or should the regulation provide merely general guidance, with numerical values provided In a regulatory guide?

ISG Response: See ISG Response to Question 1 above .

• Numerical values should not be codified in the siting regulations. Any numerical values should be placed in
• regulatory guides, as is the case under present requirements.

(b) Assuming numerical values are to be codified, are the values

• of 500 persons per square mBe at the time of site approval and 1000 persons per square mDe 40 years thereafter appropriate?

If not, what other numerical values should be codified and what is the basis for these values? ISG Response: No changes should be made to the present siting regulations. See ISG Response to Question 1 . As discussed in the ISG Response to Question 1 and the

• referenced sections of the ISG Comments, the proposed changes to the siting requirements lack an adequate technical
• basis. Any ch~nges to the regulations must have an adequate technical basis. At this time, no adequate technical basis to support any numerical values has been identified .

                                - A5 -

                                                     \

Additionally, the requirement to project population densities

• out to 40 years is problematic. Would there be safety significance if the projections were exceeded? If not, why should the projections be made in the first place? If exceeding
• the projections has safety signtflcance, what regulatory measures would NRC take?

I

 * (c) Should population density be specified out to a distance other than 30 mDes (50 km), for example, 20 miles (32 km)? If a different distance is recommended, what Is Its basis?

ISG Response: No changes should be made to the present siting regulations. See ISG Response to Question 1 . As discussed in the ISG Response to Question 1 . and the I referenced sections of* the ISG Comments, the proposed changes to the siting requirements lack an adequate technical basis. Any changes to the regulations must have an adequate technical basis . I

                           - A6 -

4. Should the Commission approve sites that exceed the proposed population values of 10 CFR 100.21, and if so, under what conditions?

• ISG Response: Yes, the Commission should approve sites that exceed the proposed population values of 10 CFR 100.21 if the
• results of an evaluation of the best available sites, considering all relevant factors, lead to selection of such a site. The Commission makes clear in the Statement of Considerations that "numerous risk I

studies on radioactive material releases to the environment under severe accident conditions have all confirmed that the present siting

• practice is expected to effectively limit risk to the public." 57 Fed .

Reg. 47,803. Moreover, as discussed in Section ll(A)(2)(b), population densities far in excess of 500 persons/square mile would

• not cause risks to exceed the safety goals. More importantly, as discussed in Section 11 (A) (1) , population is but one factor to be
• evaluated during the site selection process. Selection of sites among the *best reasonably to be found requires consideration of additional factors, such as surface faulting, seismicity, suitability of subsurface
• material, vulcanism, flooding, extreme meteorological phenomena, man-induced events, dispersion in air, dispersion in water, emergency planning, land use, availability of cooling water, and other site I

characteristics as appropriate, such as avalanche, landslide and surface collapse. It is essential to ensure that all site-related

• - A7 -

characteristics have t:,een taken into account during the selection of

• the preferred candidates. As discussed in Section ll(A)(2)(c)(iii) of the ISG Comments, the adverse impacts of the proposed revisions greatly exceed their benefits. The proposed revisions, if adopted, would
• impose a hierarchy of site characteristics, which elevates demographics over other physical characteristics of the site and other safety-related aspects of nuclear power plant siting which may have
• greater potential for reducing risk. This, in turn, would create the possibility that sites with a better balance overall of favorable safety-I related characteristics mights be eliminated from further consideration on the basis of demographics alone .
• 5. Should holders of early site permits, construction permits, and operating llcense permits be required to periodically report changes in potential offsite hazards (for example, every 5 years within 5 mHes)? If so, what regulatory purpose would such reporting I requirements serve?

ISG Response: ISG Members question why existing reporting requirements are not sufficient for the reporting of significant

• changes. If a change presents no significant hazard, why would the NRC wish to impose new reporting requirements? Wouldn't this result In increased costs without a commensurate increase in
• protection of public health and safety? If a change would potentially present a significant hazard, wouldn't the current U.S. regulations
• -AS -

require analylsis of its significance and a report of the change to the

• NRC if the change was found to present a significant hazard?

6. What continuing regulatory significance should the safety

• requirements In 10 CFR part 100 have after granting the initial operating license or combined operating llcense under 10 CFR part 527 ISG Response: See ISG Response to Question 1 above. The ISG I does not believe it is necessary to modify Part 100 and, hence, change the regulatory significance of Part 100 from what it is today .
• 7. Are there certain site meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are the conditions that can not be adequately compensated for by design features?
• ISG Response: Unfavorable meteorological conditions alone are not sufficient to reject candidate sites .

 - 8. In the description of the disposition of the recommendations of the Siting Policy Task FQrce report (NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to reconsider any

• recommendation not adopted and, if so, what are the bases for reconsideration?

ISG Response: As discussed in Section ll(A)(2)(a)(ii) of the ISG

• Comments, the Siting Policy Task Force Report (NUREG-0625) should not be used as the basis for making changes to the regulations. The Siting Policy Task Force report was issued in 1979. Since that time, I

                                  - A9 -

new information regarding severe accident phenomena, probability and consequences has been developed and new regulations established which invalidate assumptions underlying the report's recommendations. The limitations in the use of NUREG-0625 were

• recognized as early as 1979 by the Director of NRC' s Office of Standards Development and the Director of NRC's Office of Management and Program Analysis. Given regulatory developments I

since then, there are no compelling reasons to reconsider any additional recommendations of the Siting Policy Task Force Report. I B. REACTOR SITING CRITERIA {SE/SM/CJ

1. In making use of both deterministic and probabDistlc evaluations, how I should they be combined or weighted; that is, should one dominate the other?

ISG Response: As discussed in the main text of our comments (see Section ll(C)(1 )), ISG Members are not in favor of requiring the use, by regulation, of both deterministic and probabilistic methods to determine the Safe Shutdown Earthquake. By the Commission's own

• admission, the present regulation has worked reasonably well for two decades. Also, experts differ on estimates of the largest earthquakes
• and choice of ground-motion models. The Supplementary Information published with the proposed rule makes clear there is controversy over the kind of probabilistic methods to use and the I

                                  - A10 -

I

balance to be struck between probabilistic and deterministic methods.

• All of this underscores the prematurity of codifying the proposed regulations at this time where there are so many unanswered questions. The need for further evaluation by the Commission is
• reinforced by the alternative to the proposed changes to the seismic criteria submitted by the Nuclear Management and Resources Council (NUMARC) as part of its comments .

ISG Members believe that a more prudent course is to continue

• evaluation until consensus is reached on an appropriate approach.

Absent such consensus, it is highly unlikely that the proposed revisions, if adopted, will lead to anything other than regulatory

• instability.

In no circumstances should the Commission codify a requirement to use both deterministic and probabilistic analyses

 -    without prescribing a way to reconcile differences in the analyses .

Without a reconciliation method, it is certain that closure of seismic issues in the licensing process would be vastly more difficult .

2. In making use of the probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015, is [sic] the proposed

• procedures in appendix C to DG-1015, adequate to ,determine controlling earthquakes from the probabilistic analyses?

ISG Response: As part of Its comments on the proposed revisions to the seismic criteria in Part 100, the Nuclear Management and

                                - A11 -

Resources Council (NUMARC) submitted comments on DG-1015, which included a major markup of appendix C. In responding to this question, NUMARC requested that the Commission carefully evaluate NUMARC's alternative before adopting any revisions to the seismic

• criteria and implementing regulatory guides. ISG Members believe that further evaluation of the proposed revisions and alternatives
• thereto is highly desirable before the Commission adopts any changes to the seismic criteria presently in Part 100 .
• 3. The proposed Appendix B to 10 CFR part 100 has Included in Paragraph V(c) a criterion that states: *The annual probabHity of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median annual probabDity computed from the current [EFFECTIVE DATE OF THE
• FINAL RULE] populatlon of nuclear power plants." This is a relative criterion without any specific numerical value of the annual probability of exceedance because of the current status of the probabilistic seismic hazard analysis. However, this requirement assures that the design levels at new sites wUI be comparable to those at many existing sites, particularly more recently licensed sites.

Method dependent annual probabilities or target levels ~ . 1 E-4 for LLNL or 3E-5 for EPRI) are identified In the proposed regulatory guide. Sensitivity studies addressing the effects of different target probabDities are discussed in the Bernreuter to Murphy letter report.

• Comments are solicited as to: (a) whether the above criterion, as stated, needs to be Included in the regulation? and, (b) If not, should It be Included in the regulation in a different form ~ . a specific numerical value, a level other than the median annual probability computed for the current plants)?
• ISG Response: As discussed in Section II (C) of our comments, ISG Members do not believe the proposed revisions to the Commission's
• seismic regulations in Part 100 should be adopted .

                              - A12 -

4. In determining the controlling earthquakes, should be [sic] median values of the seismic hazard analysis, as described in appendix C to Draft Regulatory Guide DG-1015, be used to the exclusion of other statistical measures, such as mean or 85th percentDe? (The staff has selected probability of exceedance values associated with the median hazard analysis estimates as they provide more stable estimates of controlling earthquakes.)

• ISG Response: There is no scientific or regulatory justification for choosing the median, particularly because each existing plant has been judged to be acceptable in seismic terms. Hence, every seismic I

spectrum for existing plants should be acceptable. However, if the Commission decides to require the use of probablistic criteria, then it

• would be appropriate for the choice of controlling earthquakes to employ the Safety Goal decisional framework. That ls, for very severe, very rare natural events, a useful criterion might be that for
• such events, the incremental harm due to the presence of the nuclear plant be small compared to that due to the event itself.

5. For the probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power plants? (For the four trial plants used to develop the

• criteria presented In Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probabDlty of exceedance level. Controlling earthquakes were evaluated for this frequency band, for the average of 1 and 2.5 Hz spectral responses, and for peak ground acceleration.)

I ISG Response: See above response to Question 2 (seismic).

                              - A13 -

I

                                           ?    ,~~~LNDU~~~;    PR  S0 (5'7Ff2.. ~1fc}_)

NORMAN R. TILFORD Consulting Geologist Environmental/Engineering Geology P.O. Box 1119 *93 JU -1 Pl ? :16 Hilltop Lakes, TX 77871 Office: Texas A & M University: 409 845-9682 Fax: '409-845-6162 May 27, 1993 Secretary, USNRC Attn. Docketing & Service Br. Washington, DC 20555 Gentlemen: The attached material was developed in response to a client request for review of the proposed changes by the NRC to 10 CFR 100, Appendix A. The client has subsequently allowed me to forward them to you as a comment to NRC. Thank you for your consideration and the inclusion of these comments in the record.

• iZf.P Norman R. Til d JUL 3 O1993 Acknowledged by card .......................*-**-*

         - ... ":*MW,~S!O,
                 *nuN AR l,          3

1 Comment on Proposed Rule Changes to 10 CFR, Part 100 Appendix A-B by Norman R. Tilford Center for Engineering Geosciences Department of Geology Texas A & M University College Station, Texas May 6, 1993 1 Civilian nuclear power plant site selection studies and decisions relative to geologic and 2 seismologic factors have been guided by Appendix A to 10 CFRlOO for about 20 years. Appendix 3 A has served the interest of public safety well. There is no documented instance of release of 4 radioactive materials from any commercial nuclear reactor in the United States as a result of *, 5 unsatisfactory performance of a nuclear site from geologic or seismic factors. One of the most 6 interesting and revealing bases for the "new" probabilistic approach to selecting sites for nuclear 7 power plants, is the proposed reliance on sites and design values established under Appendix A. 8 *This forms the highest possible form of commendation for the existing regulation. 10 Geological studies performed as part of the site selection and licensing pn;>cess have produced vital 11- W:onnation used to establish design and operating parameters. Geologic studies offer information, 12 insight, and a perspective that cannot be obtained in any other way. Ultimately, all of the opinions ~~ formulated by experts are based on facts and information flowing from geological data. Thus the opinion of experts, at best, is only as good as the information that is available upon which the 15 opinions are based. 16 17 There are four parties at interest in the regulation of site selection for nuclear power facilities: 18

• the Public.

19

• the Regulator.

20

• the Regulated Industry or Proponent. .

21

• the Scientific and Technical Communities.

23 The divergence in belief systems, methods, and format and content of results within the scientific 24 and technical communities create broad and deep disagreement with respect to attainment of 25 acceptable levels of public health and safety in the arena of geologic hazards. It is corrnnon for 26 adherents of the probabilistic approach to describe natural observational science as "detenninistic" 27 and to suggest a lack of flexibility in that approach in the past. Such descriptions cannot be 28 accepted uncritically. In fact the deterministic approach from natural science has always 29 incorporated probabilistic methods. Certainly, this has been true during the twenty odd years of 30 nuclear power plant site selection and licensing. Actually, probability is a part, or use, of 31 determinism. Hume (1777) offers the following philosophical underpinning on which much of 32 mcxlem science and engineering rest: -

2 34 "We have said that all arguments concerning existence are founded on the 35 relation of cause and effect; that our knowledge of that relation is derived 36 entirely from experience; and that all our experimental conclusions 37 proceed upon the supposition that the future will be conformable to the 38 past.----" 39 "Nothing so like as eggs; yet no one, on account of this appearing 40 similarity, expects the same taste and relish in all of them. It is only 41 after a long course of uniform experiments in any kind, that we attain a 42 firm reliance and security with regard to a particular event.----" 43 "If we be, therefore, engaged by arguments to put trust i'n past experience, ~4 and make it the standard of our future judgement, these arguments must. 45 be probable only-----" (my underlining) 47 A recently popular saying has it that, If it looks like a duck, sounds like a duck, and walks like a 48 duck, then it is a duck." This is both a deterministic statement from experience (experiment) and a 49 statement of likelyhood, or probability. The confidence with which the speaker makes the above 50 statement is simply a statement about the probability that a class of objects is what it seems. 51 Experience tells us that the probability of the correctness of this statement is acceptably high. 52 Lacking a basis in experience, no such statement could be made, or supported, or accepted. Also, 53 absence of such a basis in experience would leave no justification for assessing the probability of 54 the truth of the statement 55 56 Regulation in the public interest is synonymous with concern for public health and safety. The 57 .6 growth of opposition of the public to site selection decisions for nuclear and other industrial 58 facilities in the past two decades has been remarkable. This can be taken as a statement on the part of the public that the regulatory process intended to protect public health and safety has been corrupted. 62 The massive public rejection of regulatory decisions in this arena can be met in only two ways; 63 education, or coercion. Education, along with some reform of the relationship among the 64 regulated, the regulator, and the scientific and technical community, is the best response. Such a 65 response is necessary and must not be set aside in deference to the establishment of new regulatory 66 relationships. Tolerance of some uncertainty and ambiguity is a necessary condition of human 67 existence, and the existence of human institutions. 70 The relationship between a regulated industry or proponent and a regulatory authority is always 71 characterized by pressures to move toward prescriptive regulation rather than performance 72 , regulation. The regulated industry would always prefer to decrease uncertainty with respect to the 73 regulatory outcome and would always prefer to reduce the expense, capital risk, and time 74 investment necessary to achieve regulatory approval. The regulatory agency and staff are therefore 75 l,IIlder constant pressure to develop means to reduce the need for quasi-subjective judgments and 76 substitute requirements which lead to a check-off approach. These forces seldom, if ever, have 77 anything directly to do with the satisfaction of need for public health and safety, which is the basic 78 justification for the regulatory process.

3 80 The goal of increasing "stability" in the licensing process as proposed by the nuclear industry is not 81 satisfying. Physically any process which continues without change is a stable process. Celestial 82 bodies are st.able in their orbit Flows of rivers or streams through annual, decadinal and other 83 time-based periods can be characterized as stable, although within that stability, there is an 84 enormous capacity for destruction of human edifices and life. The same can be said to be true of 85 earthquakes, hurricanes, tsunami, and many other naturally occurring phenomena. There is, 86 therefore, no inherent benefit to public health and safety in the concept of stability with respect to 87 licensing. There is, in fact, no established relation between stability and safety. 89 Much of the progress in our knowledge of the earth's crust during the nuclear development period 90 since the late 19(i()'s rests on work done to develop safe design bases for these plants. The 91 industry solution would freeze the status quo and eliminate incentives which justify basic research. 92 The representatives of the nuclear power industry would have the NRC accept the use of

• i93 96 97 98 probability based on expert opinion, and the design bases of existing nuclear plants, as the default basis for future site selection and evaluation. This present attempt of the nuclear industry to codify the status quo in the selection of nuclear sites is unacceptable precisely because stability in licensing cannot be assured or demonstrated to serve the public need for protection of health and safety.

Since the data base on which any present statements of probability rest is so scant, the first large earthquake in the eastern half of the country which occurs in the future will radically alter our base 99 of experience. If we freeze the regulatory basis and methodology on present knowledge, the entire 100 framework Ell.be altered by a single event 102 This was recogniz.ed clearly by the National Research Council (1988) panel in their publication on 103 Probabilistic Seismic Hazard Analysis, page 6(Uil: "In lower actlv:l.ty environments, 104 such as the eastern United States, ground motion results from PSHA and 105 deterministic methods may be very different, even at low probability levels, 106 because of the lower recurrence rates in these regions. At low probability 107 levels, b.2th deterministic and probabilistic hazard studies should be 108 conducted to arrive at appropriate seismic design or evaluation criteria. ---At -~ 109 113 114 present the uncertainty regarding the seismic hazard at certain locations in the United States is highly variable. To a large extent this stems from variability in the knowledge of the sources and rates of seismic activity." The recent highly accelerated growth of knowledge about geological hazards in relation to mankind has been driven materially by the need for public health and safety information in the nuclear power 115 industry and it's regulatory process. Pressures to abandon the natural science observational 116 approach to study of earthquakes and other geological haz.ard can be compared to pressures to 117 abandon the search for medical cures for difficult diseases, such as cancer. If we abandon the 118 search for cures for cancer on the basis that the probability of an individual catching the disease is 119 small and acceptable, then the public would reject the idea as anti-humanistic, mechanical, and 120 unresponsive to the need for public health and safety. So it is when we attempt to put aside 121 uncertainty and the process of the growth of knowledge spurred by the nuclear power industry. 122 While the need for realistic design bases for nuclear and other critical industrial facilities is clearly 123 accepted, it is irresponsibile to foster the perception of eJiroioation of uncertainty through the use of 124 expert opinion and probability. All studies we know of using expert opinion show the wide range 125 of opinion which can be developed using available information. The adoption of such a single-126 minded approach would lead to further erosion of public confidence in the regulatory process. 128 Within the scientific and technical community there are basic divisions of approach and interest, 129 making up three groups. One leg or point of this triangle is that of the natural science community,

4 130 in which the observation of natural phenomena, the design of experiments to elucidate the cause 131 and effect relationships between these phenomena, and the development of hypotheses leading to 132 theory and establishment of natural law, are the guiding paradigm. 134 A second cormmmity is that group involved with the manipulation of the database created from 135 activities of the natural science observational and experimental group. This community includes 136 classical mathematical seismologists and mathematical modellers of various persuasions. In 137 regulation, the activities of this community are characterized by the manipulation of an existing 138 database to produce results which satisfy the need to reach conclusions in decision making 139 processes. 140 141 The third community is the engineering community. This community looks to the two other 142 groups to provide inputs that serve as the basis for the design of engineered structures which -4 143 146 147 adhere to the laws of physical forces as known and understood by the engineering connnunity. Modellin bawfty) The natural science community relies on Hume's (1777) statements about the 148 149

    +I                                                      Uniformity of Nature, which conclude that the future will resemble the past Hutton (1795) 150  I

• further elucidated this with respect to geologic I

151 I phenomena in his Concept of Uniformity, 152 I which states that the earth we see is the result of 153 I forces acting in the past which are similar to 154 I I those we see acting today. Gould, (1987) has . 155 embraced the growing body of evidence that 156 geological uniformity includes punctuating 157 catastrophic events. All these insights rely on 158 F.ngineering . the relation of cause and effect in our world, 159 ~ ) --------+ (Implementation) and explicitly embrace the idea that causes can -K~ be deduced from effects. The natural science community faces difficulties in identifying, defining and codifying natural laws 163 to enable prediction of earthquakes and other naturally occurring phenomena. The most crucial 164 problem is the time-dependent nature of full experiental disclosure of the nature of physical 165 processes acting in the earth environment Full experience of the parameters influencing or 166 controlling the occuttence of destructive earthquakes may be dependent upon observations which 167 would extend over thousands of years, and possibly tens of thousands or hundreds of thousands 168 ~~~- ' 169 170 The interaction of the natural science, modelling, and engineering groups has created pressmes to 171 abandon or decrease the use of natural science observational and experimental input in favor of a 172 systematic industrial attempt to make decisions with available technology, while effectively 173 freezing the status quo through probabilistic approaches and modelling. Modellers are constrained 174 by shortcomings of the database furnished to them by the natural science observational community. 175 The models produced can only be as good as the poorest input information. Garbage in. ... garbage 176 out! I Therefore they attempt to overcome the problem by addressing elements of uncertainty in 177 assumptions and other modifying elements of model formulations. In the past few years, through 178 the application of Delfi and other management and psychologically based techniques, the modelling 179 community has incorporated the use of expert opinion to modify the perceived statistical basis for 180 decision making. See, for example, National Research Council, (1988).

5 182 Reiter (1990) identifies two somces of systematic uncertainty in statistical manipulation of the 183 sparse data set available: scientific uncertainty and informational uncertainty. "One of the main 184 problems associated with the systematic uncertainty in seismic hazard estimation is that it is 185 determined in large by the use of expert judgement.... Large differences of opinion exist and 186 although great efforts have been made to reduce these differences (see for example, Electric Power 187 Research Institute 1986), to a large extent, they still remain." 188 189 Meanwhile, the natural science community is utilizing evolving and often indirect means of 190 observation and experiment to move our understanding of natural hazards forward in quantwn 191 leaps relative to the length of time needed to complete a full cycle of direct experience. Under 192 Appendix A to 10 CFR 100, and through the explosive evolution of technological observation -5 193 capabilities, a conservative statement would be that geologists, seismologists, and engineers have 194 learned more about earthquakes and the crust of the earth in the past 50 years than in the preceding 500 years. 197 A t,a9lc tenet of earth 198 eclence, corollary to the le.lea 199 that the present Is the key to 200 the past and future, le the 201 Idea that most th Inge that 202 have happened, and will 203 Earthquake happen on the earth, prolnlt,ty 204 s1z.es .Am happening at some point 205 on the earth In modem 206 tlm85. We thus look for 207 present analogs to 208 understand conditions and 209 evente we see In the geologic reGOrd. In this view. WB can j~ 212 duplicate ol7servlng a certain area on the earth for a 213 214

                             ~ ...----------------prolonged period t,y looking Timetocomplctecycleofcarthquakcs          at the entire globe for a 215                                                                            short period.
         *- Typical earthquake 217                                                                            Those opposed to research by 218 private industry in support of public safety state that the unknowns surrounding naturally occurring 219 phenomena are not the responsibility of a segment of industry seeking licensing approval through 220 the regulatory process. While this argument may be seductive it is misleading. In our society it' 221 should be the responsibility of those seeking profit from a particular activity to provide adequate 222 assurances of the protection of public health and safety within the American system. It is therefore 223 unacceptable and inappropriate for such institutions-for-profit to argue that it is the responsibility of 224 others to carry out the research necessary to establish appropriate physical parameters which may 225 govern public safety, resulting from their own industrial activities. The responsibility for 226 satisfying the need for public health and safety lies, in fact, with proponents of the integration of 227 new technologies into the Earth system.

229 Statements by the Numarc Ad Hoc Advisory Committee on the draft regulatory guide DG-1015,

6 230 "Identification and Characterization of Seismic Sources and SSE Ground Motion", imply the 231 existence of a factual basis for the estimates prepared by the Electric Power Research Institute 232 (EPRI) and Lawrence Livermore National Laboratory (LLNL) relative to the probabilistic basis for 233 determination of safe shut-down earthquakes (SSE). This implication of the existence of a factual 234 scientific basis for the conclusions identified by these studies (as facts seeking confinnation) 235 constitute presumptive, coercive and misleading propaganda with respect to public acceptance of 236 such a regulatory methodology. 239 240 SEPARATE ISSUE: 242 Federal Register Volume 57 Number 203 Dated Tuesday, October 20, 1992 on the topic of 243 "Nuclear Regulatory Commission 10 CFR Parts 50, 52, and 100, Subpart 5 Major Changes A. 4f4 Reactor Siting Criteria (Non-seismic); PartXI Questions" 246 This comment is addressed to Question 7 on page 47812. That question is "Are there certain site 247 meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are 248 the conditions that cannot be adequately compensated for by design featuresT' 250 RESPONSE: Yes. A significant condition which cannot be adequately compensated for by 251 design features would be the elimination of the site by erosion from flooding and other extreme 252 meteorological or geological events such as hmricanes or major coastal subsidence. There is 253 substantial geologic evidence of major hurricanes in the recent geological past which have radically 254 altered the coastal environment of significant parts of the eastern U.S. seaboard and Gulf Coast 255 region. Sites being proposed at these locations should be scrutiniz.ed carefully for safety against 256 destruction of the site caused by extreme external events. ,. tI REFERENCES NATIONAL RESEARCH COUNCIL, 1988, Probabilistic Seismic Hazard Analysis: National 263 Academy Press, Washington, DC 265 REITER, LEON, 1990, Earthquake Hazard Analysis: Issues and Insights: Columbia University 266 Press, New York, NY 268 HUME, DAYID, 1777, An Enquiry concerning the Human Understanding. and an Enquiry 269 concerning the Principles of Morals: Oxford University Press, Oxford, UK 271 GOULD, S.J., 1987, Times Arrow, Tnnes Cycle: Harvard University Press, Cambridge, MA 273 HUTTON, JAMES, 1795, Theory of the Earth, with Proofs and illustrations: W. Creech, 274 Edinburgh,Scotland

DOCKET NUMBER

• PROPOSED RULE PR 50 2J-J O/J Department of Energy (Sl ff< L/ 1 Washington, DC 20585 (': _ * :.

                                                                                'J May 28 , 1993 t.,r .I:...

Mr. Samuel Chil k Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 ATTN: Docketing and Service Branch

Dear Mr. Chilk:

This letter provides comments from the Department of Energy (DOE), Office of Nuclear Energy, on the proposed revisions to Parts 50, 52, and 100 of Title 10 of the Code of Federal Regulations (10 CFR), published in the Federal Register, 57 FR 48712. We note that the revised Part 100 replaces the determination of the Exclusion Area Boundary {EAB) distance with an arbitrary selected value of 0.4 miles, and substitutes present and future population density criteria for the determination of a Low Population Zone {LPZ). The selected value for the exclusion area distance would exclude a number of existing sites, if future plants were to be sited on them. In light of the expectation that future plants most likely will be Advanced Light Water Reactors (ALWRs), and that ALWRs have improved safety characteristics as well as severe accident risk profiles an order of magnitude lower than existing plants, this EAB criterion sends an incorrect and confusing signal to the public. Plants with improved safety characteristics should not require greater exclusion areas than operating plants, which have been found safe by the Nuclear Regulatory Commission (NRC). We recommend that the value selected as the minimum EAB distance be selected to be compatible with the minimum EAB found to be adequate by NRC for operating plants. With respect to the proposed population density criteria, we have performed two background studies of the concept of using population density as a method for judging the potential impact of the plant on the population in the vicinity of the pl ant. JUL 3 0 1993 Acknowledged by card ........................... n.,.: 1

~1.S '- L':.: 1..,*,

2 The first of these studies was conducted by the joint industry-DOE sponsored Early Site Permit program (1), and a second independent study by Sandia National Laboratories (2). Both studies conclude that uniform population density is not a useful parameter for judging the impact of the pl ant on population in the vicinity of the site. From these studi es we conlude that t he existing concept of a LPZ, as defined in Part 100, provides a better approach for factoring nearby population centers into siti ng decisions, and avoiding sites in proxi mity to high populati on densities, than the proposed uniform population density criteria. We recommend, therefore, that the population density criteria in the proposed revisions be deleted, and that the requirements for defining a LPZ surrounding the plant be retained in Part 100. In summary, while we fully endorse the NRC intent to update the assumptions for the source terms and dose calculations, as noted in the statement of consideration in the proposed rulemaking announcement, we recommend that the criteria for future site selections not be any more restrictive than the current criteria. We suggest that this can be accomplished by selecting a minimum exclusion area boundary of 0.25 miles, and keeping the concept of a LPZ, as presently defined in Part 100. Sincerely,

                               <1 :>~JrrJd.p,j
                             ., E. C. Bro l in, Act ing
                             ,   Assistant Secretary for Nuclear Energy (1) J . Ziegler, "Evaluation of Population Distribution Relative to Meeting the Quantitative Health Obj ective of the NRC Safety Goal for Off-Site Risk Associated with Nuclear Power Plants," NUS Corp. Report, Early Site Permit Demonstration Program, March 1993.

(2) M. Young, "Evaluation of Population Density and Distribution Criteria in Nuclear Power Plant Siting," SAND93-0848, Sandia National Laboratories, in preparation.

1t,IJ NUCLEAR MANAGEMENT AND RESOURCES COUNCIL 1776 Eye Street, NW.

• Suite 300
• Washington. DC 20006-3706 (202)872-1280 WIiiiam H. Resin Vice President & Director Technical Division May 28, 1993 Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

- Attention: Docketing and Service Branch

Dear Mr. Chilk:

The Nuclear Management and Resources Council (NUMARC) 1 had submitted a comment package in response to the proposed rule, 10 CFR Parts 50, 52, and 100, Reactor Siting Criteria, (Federal Register 47802, October 20, 1992, and 55601, November 25, 1992) on March 24, 1993. This letter and its enclosures are additional comments being submitted for consideration. In our March 24, 1993, comment package, we recommended modification of the Safe Shutdown Earthquake Ground Motion (SSE) sections in 10 CFR Part 50, Appendix Sand in Standard Review Plan (SRP) Section 2.5.2. The recommended language would permit usage of future technological advances in determining the effective response of nuclear power plant structures and systems to high frequency motions. In order to provide consistency between the draft regulatory guides and 10 CFR Part 50, Appendix Sand SRP Section 2.5.2 with respect to the aforementioned modifications, we are submitting a revision to Appendix F to Draft Regulatory Guide 1NUMARC is the organization of the nuclear power indusny that coordinates the combined efforts of all utilities licensed by the NRC to construct or operate a nuclear power plant, and of other nuclear industry organizations, in all matters involving generic regulatory policy and on the regulatory aspects of generic operational and technical issues that affect the nuclear power industry. Every utility responsible for constructing and operating a commercial nuclear facility is a member of the NUMARC. In addition, NUMARC's member include major architect-engineering firms and all the major steam supply vendors.

Mr. Samuel J. Chilk May 28, 1993 Page2 DG-1015 proposed by industry as part of our earlier comment package. Enclosure 1 provides the revised Appendix F with the most recent changes highlighted in bold print. Enclosure 2 is a draft ofEPRI Report TR-102470 entitled, "Analysis ofHigh-Frequency Seismic Effects." It provides the technical basis for the modifications recommended in the March 24, 1993, NUMARC comment package and in the enclosed revised DG-1015, Appendix F. We believe that implementation of this technological advance by industry will reduce costs related to design and construction of future nuclear power plants without adversely affecting public health and safety. We would welcome further, public dialogue with the staff to explain our comments or to answer any questions related to the enclosed EPRI report. Sincerely, ti~ William H. Rasin NMF/ljw Enclosures c: Mr. James M. Taylor, EDO, NRC (w/o enclosures) Mr. Eric S. Beckjord, Director, RES (w/o enclosures) Dr. Thomas E. Murley, Director, NRR (w/o enclosures) Dr. Andrew J. Murphy, RES (w/ enclosure) Mr. Gautam Bagchi, NRR (w/ enclosure)

                                                                                   ~
                                                                                    ./

Mr. SamuelJ. Chilk May 28, 1993 Page3 be: D. Modeen (w/o enclosure) N. Farukhi J. Eckart, Newman & Holtzinger J. Devine, EPRI ALWR Program B. Pusheck, EPRIALWRProgram J.C. Stepp, EPRI RP. Kassawara, EPRI B. McIntyre, Westinghouse C. Brinkman, ABB/CE J. Quirk, GE W. Pasedag, DOE G. Vine, EPRI (DC) File: 15.4.2

Enclosure 1 NUMARC Comments May28, 1993 Line lnllil8 out 1 APPENDIXF 2 3 PROCEDURE TO DETERMINE THE SAFE SHUTDOWN EARTHQUAKE GROUND 4 RESPONSE SPECTRUM 5 6 1 F, 1 Introduction 8 9 This appendix describes an acceptable procedure to determine the safe shutdown __10 earthquake (SSE) ground response spectrum. The ground response spectrum is defined in terms 11 of the horizontal and vertical motion at the free-field ground surface at the plant site. It is 12 developed with the consideration of site seismic wave transmission effects and the mean 13 magnitude and distance of earthquakes that produce the SSE (see Appendix C of this Regulatory 14 Guide). 15 16 The SSE for a site is determined by the procedure described in Appendix B to this 17 Regulatory Guide for the average spectral acceleration between 5-10 Hz and 1-2.5 Hz. Two 18 frequency ranges are considered to insure that a wide range of frequencies are considered in the 19 development of the SSE response spectrum. The procedure in Appendix B determines the SSE 20 that is consistent with the design level for existing plants as of [Effective Date of the Final Rule]. The SSE ground response spectrum is determined by scaling a spectral shape to the levels for 5- ~~ 10 Hz and 1-2.5 Hz. Based on effective acceleration criteria (Ref. SF), amplitudes of ground 23 motion frequencies greater than 10 Hz may be reduced for the site-specific response 24 spectrum. It is anticipated that a future Regulatory Guide will provide guidance in the 25 assessment of site-specific response spectra 26 27 F.2 Procedure 28 29 The SSE ground response spectrum is detennined,by scaling an 84th percentile response 3o spectrum shape to the average spectral acceleration lev~ls corresponding to the Reference 31 Probability as defined in Appendix B to this Regulatory Guide. The steps in the procedure are: F-1

NUMARC Comments May28, 1993 Line ln/Une out 1 Step 1. Detennine the average spectral acceleration level for 5-10 Hz and 1-2.5 Hz from 2 the median hazard curves at the Reference Probability (see Appendix B). 3 4 Step 2. Determine an 84t h percentile response spectrum shape for the site. A site-specific 5 or standard response spectrum shape can be used (Ref. IF). Two response 6 spectrum shapes are required; one that is scaled to the 5-10 Hz (7.5 Hz) average 7 spectral acceleration and another which is scaled to the 1-2.5 Hz level (1. 75 Hz). 8 Subsection F.3 identifies acceptable procedures to determine the response 9 spectrum shape. 10 11 A standard response spectrum refers to a ground response spectrum that is 12 independent of earthquake magnitude and distance and that accounts for site 13 conditions in terms of general site categories (e.g., rock or soil). 14 15 Step 3. The response spectrum shapes determined in Step 2 are scaled to the 5-10 Hz and 16 1-2.5 Hz average SSE levels, respectively from Step 1. This step is illustrated in 17 Figure F.l. 18 19 Step 4. Step 3 produces two response spectra, one scaled to the average spectral .0 acceleration between 5-10 Hz and the other between 1-2.5 Hz. For purposes of 21 the site characterization, the applicant can envelope the two spectra or 22 alternatively, elect to retain two spectra that are considered in design evaluations. 23 The latter approach may be preferred when the response spectra have different 24 spectral shapes. This is illustrated in Figure F.2. 25 26 F.3 Ground Response Spectrum Shape 27 28 The response spectrum shape for a site should be developed considering the response of 29 surficial soil deposits to earthquake ground motion and the mean magnitude and distance of 30 earthquakes that produce the SSE. Alternatively, a standard response spectrum shape of the type F-2

NUMARC Commerta May28, 1993 Line ln/Ltne oii 1 used in past nuclear power plant designs is acceptable (Ref IF). Currently, the Electric Power 2 Research Institute (EPRI) is completing work on the development of a ground motion model for 3 sites (rock and soil) in the stable continental region (SCR) (Ref. 7F). It is anticipated that this 4 work will be incorporated in a future Regulatory Guide that provides specific guidance in the 5 assessment of site-specific ground motion in the SCR. 6 7 Site-Specific Response Spectrum 8 9 The development of a site-specific response spectrum shape should consider the site soil 10 and foundation properties, regional seismic wave propagation and the mean magnitude and 11 distance of earthquakes for the SSE. A response spectrum shape is determined for the mean 12 magnitude and distance for each frequency (i.e., 5-10 Hz and 1-2.5 Hz). Ground motion 13 amplitudes at frequencies greater than 10 Hz may be appropriately reduced after due 14 consideration has been given to the criteria on the effectiveness of motions at high 15 frequencies (Ref. SF). The procedure to determine the mean magnitude and distance is 16 described in Appendix C of this Regulatory Guide. 17 18 Methods to determine a site-specific response spectrum include: 19 ~o 1. Development of a database of strong motion records that are selected to have 21 magnitude and distance characteristics similar to the mean magnitude and distance 22 for the SSE. In addition the strong motion records should have similar earthquake 23 source characteristics, propagation path, and recording site properties. From the 24 database of strong motion records an empirical response spectrum shape can be 25 developed (Ref. 2F). The 84th percentile response spectrum shape is used in the 26 assessment of the SSE ground response spectrum. While this approach may be 27 preferred for some sites, it does not explicitly account for randomness and 28 uncertainty in ground motion. 29 30 2. A response spectrum shape can be developed using theoretical-empirical modeling F-3

NUMARC Comments May28, 1993 Lila ln/Une Out 1 techniques. These methods can be used to model conditions that are not well 2 represented in the strong motion record database and to fully model randomness 3 and uncertainty. The EPRI ground motion model for the SCR is such an example 4 (Ref. 7F). The 84th percentile response spectrum shape may be used in the 5 assessment of the SSE ground response spectrum. 6 7 3. Analytical methods can be used to model the response of local site soil conditions 8 (Ref 3F-7F) to earthquake ground motions. The response of the site soils should 9 be evaluated for earthquake motions defined at free-field ground surface on rock 10 associated with the mean magnitude and distance determined for the two 11 frequency ranges considered (e.g., 5-10 Hz and 1-2.5 Hz). 12 13 Standard Response Spectrum Shape 14 15 It is acceptable to use a standard response spectrum shape to detennine the SSE. 16 However, since existing shapes were normalized to the peak ground acceleration, they should be 17 regenerated based on a scaling to the average spectral acceleration for 5-10 Hz and 1-2.5 Hz. F-4

NUMARC Corrrneuts May28, 1993 Line In/line Out 1 REFERENCES 2 3 IF. U.S. Nuclear Regulatory Commission, Design Response Spectra For Seismic Design For 4 Nuclear Power Plants," Regulatory Guide 1.60. 5 6 2F. U.S. Nuclear Regulatory Commission, "Development of Site-Specific Response Spectra," 7 NUREG/CR-4861, 1980. 8 9 3F. P. B. SchnabeL J. Lysmer, and H.B. Seed, "SHAKE-A Computer Program for 10 Earthquake Response Analysis of Horizontally Layered Sites," Report No. EERC 72-12, 11 Earthquake Engineering Research Center, University of California, Berkeley, 1972. 12 13 4F. I. V. Constantopoulos, "Amplification Studies for a Nonlinear Hysteretic Soil ModeL" 14 Report No. R73-46, Department of Civil Engineering, Massachusetts Institute of 15 Technology, 1973. 16 17 SF. V. L. Streeter, E. B. Wylie, and F. E. Richart, "Soil Motion Computation by 18 Characteristics Methods," Proceedings American Society of Civil Engineers, Journal of 19 the Geotechnical Engineering Division, Vol. 100, pp. 247-263, 1974. e20 21 6F. W. B. Joyner and AT. F. Chen, "Calculations of Nonlinear Ground Response in 22 Earthquakes," Bulletin Seismological Society of America, Vol. 65, pp. 1315-1336, 1975. 23 24 7F. Electric Power Research Institute, "Guidelines for Determining Design Basis Ground 25 Motions," EPRI Report TR-102293, Vols. 1-4, May 1993. 26 27 SF. Electric Power Research Institute, "Analysis of High-Frequency Seismic Effects," 28 EPRI Report TR-102470, September 1993. F-5

NUMARC Comments Mly28, 1Qg3 Une lnlUne out D Average Spectral Accelerations Corresponding to the Reference Probability

                - - s. (RP)s-10
                                                    / ___ _J.. _ _ _
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                                          . I               I 12
                                        ./ I                I 13                                     /      I              I 14                                 /          I              I 15                                            I              I I              I 16                                            I              I 17                                            I              I 18 19 1.0    1.75     5.0   7.5     10.0 Frequency (Hz) 22 23 24 Figure F.1 illustration of the procedure to scale a response spectrum shape to the 5-10 25 Hz and 1-2.5 Hz average spectral acceleration levels corresponding to the Reference 26 Probability.

F-6

NUMARC Comments May28, 1993 Llnelnll.lneOut 1 2 Option 1 - Envelope 3 4 5 C c:: 6 0 0 7 8 ~

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A 10 11 12

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8c.. 14 c.. Cl) C/) 15 16 1.75 7.5 7.5 17 18 Frequency (Hz) Frequency (Hz) 19 20 21 22 Option 2 - Retain 2 Spectra 23 24 -26 C 25 0 ct, 27 28 a3 Q.) (,) _,,'( 29 (,)

                                              / /

30 31 32 ca (,) Q) 33 0. 34 Cl) 35 36 1.75 7.5 37 Frequency (Hz) 38 39 Figure F .2 Illustration of the options to determine the SSE response spectrum. F-7

Enclosure 2 ANALYSIS OF HIGH-FREQUENCY SEISMIC EFFECTS TR-102470 Research Project 2722-23 Prepared by John W. Reed Robert P. Kennedy Jack R. Benjamin and Associates, Inc. RPK Structural Mechan1cs Consulting 444 Castro Street, Suite 501 18971 Villa Terrace Mountain View, CA 94041 Yorba Linda, CA 92686 Bah111an Lashkari Jack R. Benjam1n and Assoc1ates, Inc. 444 Castro Street, Suite 501 Mountain View, California 94041 Prepared for Electric Power Research Institute 3412 Hillview Avenue Palo Alto, Californ1a 94304 EPRI Project Manager R.P. Kassawara Nuclear Seismic Risk Program Nuclear Power Division

DISCLAIMER OF WARRANTIES ANO LIMITATION OF LIAflILITIES This report was pr~pared by the Organization(s) named below as 1n account of work sponsored by the Electric Power Research Institute. Inc. (EPRI). Neither EPRI. any member of EPRI, the Organ,zat1on(s) named ~low, nor any person acting on behalf of any of them: (a) Makes any warranty or representation whatsoever, express or implied, (I) with respect to the use of any infomation, apparatus, method, process, or similar itea disclosed 1n this report, including merchantability and fitness for a particular purpose (II) that such use does not infringe on or interfere with privately owned* rights, including any party's intellectual property, or (III) th~t this report is suitable to any particular user's circumstance; or (b) assumes re'spons1b1ltty for any damages or other liability whatsoever (including any consequential damages, even if EPRl or any EPRI representative has been advised of the possibility of such damage(s) resulting from1your selection or use with respect to the use of this report or any fnformation, apparatus, method, process or similar item disclosed in this report. Organizat1on(s) that prepared this report: Jack R. Benjamin and Associates, Inc. RPK Structural Mechanics Consulting i1

ABSTRACT This report presents the results of a project conducted for the Electric Power Research Institute (EPRI) to investigate the potential for damage fr0111 high-frequency seismic ground motions on nuclear power plant (NPP) components. The study was performed to develop a rational procedure for reducing input to high-frequency components that are being evaluated for earthquake response spectra rich in high-frequency energy. The final results from the study are intended for use

 ,n the Individual Plant Examination for External Events (IPEEE) to resolve NRC Severe Accident Policy (SAP) issues. In addition, the results will be useful for reducing site-specific design response spectra with significant high-frequency energy.

High-frequency seismic motions (i.e., greater than 10 Hz) were investigated and found to be significantly less damaging than low-frequency ground motions. It ts concluded that structures and equipment at nuclear power plants have additional capacity above yield to absorb the small ~isplacements associated with high-frequency earthquake input. The lack of ductile capacity of limiting el&11ents when subjected to low-frequency seismic motions provides a rational basis to reduce ground response spectra at high frequencies. By maintaining a consistent margin between the failure level and the design capacity across all dynamic frequencies reduction factors are obtained. The simplified procedures that were developed can be used to reduce earthquake ground response spectra 1n the high-frequency region. The different types of structures and equipment used at nuclear power plants were reviewed and it was concluded that a conservative case is I model that fails at the very small displacement of 0,01 inch. This case corresponds to an electrical cabinet that is anchored at its base by a 3/16-lnch fillet weld loaded in the transverse direction. Two models that represent this conservative limiting case were studied where the inelastic response was concentrated in the connection between the model and the base support. A slid'ing model that considered friction between the base and the support was analyzed. It was found that the effects of friction are not significantly different from the anchorage since the anchorage yield and friction ii 1

forces are interchangeable. A rocking model that included the restoring force of gravity was also analyzed. However, it was found that the sliding model is generally more conservative. Nonlinear time history analyses of both models were conducted using 15 different earthquake records for six model frequencies (i.e., 2 to 25 Hz). ,A wide range of se1sic input that represents high-frequency, low-frequency and broad-banded motions were applied to the models. The results of these analyses demonstrate that high-frequency motions are less damaging collll)ared to low-frequency motions. The increase in capacity increases at higher and higher frequencies. Two simplified analysis procedures were developed (one for each ll!Odel) us1ng a pseudo linear-elastic approach, that avoids having to perform nonlinear time history analysis. The purpose for developing these procedures was to investigate the influence of different model parameters and to provide an easy-to-use method for reducing ground response spectra in practical applications. The simplified procedures were calibrated with the results from the nonlinear time history analyses. It was found that the same values for the two frequency and duiping el'IIJ).irical parameters can be used for both the sliding and rocking models. The differences between the simplified procedures and nonlinear results were found to be small (i.e., coefficients of variation of 0.11 and 0.16 for the sliding and rocking models, respectively). The resulting parameter values were found to compare closely with results fro111 put pseudo linear-elastic analyses for different models and from different investigators. The ground response spectrum reduction factors are simply equal to 1/F,, whtre F~ is the inelastic energy absorption factor as given tn the report. It is recoanended that F, be based on the sliding model with a per11issible anchorage distortion of 0.01 inch. Another important reduction for high-frequency input that is based on ground motion incoherence should be applied first before the inelastic energy absorption factor reduction is _applied. The reduction for ground motion incoherence is permitted because seismic-induced inotions of a large massive structure founded on a substantial size basemat are reduced from the free-field motions. Currently a test program sponsored by EPRI is underway to demonstrate the physical behavior of sliding and rocking models secured by smll fillet welds when 1v

subjected to earthquake motions. The nonlinear time-history analysis computer program that was used to develop the simplified approach in this report is also being benchmarked against the test results

• V

ACKNOWLEDGEMtNTS This report was reviewed by an EPRI panel which included Dr. C. Allin Cornell, Mr. Paul Ibanez and Or. John Stevenson. Dr. Martin W. Mccann, Jr. assisted in the selection of the time histories used in the study. vi

CONTENTS Sect1on £.m ABSTRACT iii 1 INTRODUCTION 1-1 Objective of Study 1-1 Background on High-Frequency Seismic Motions 1-1 Overview of High-Frequency Components and Basis for Study 1-8 Spat1al Variation and Incoherence of Ground Motion 1-11 Organization of Report 1-15 2 BASIS, FINDINGS ANO RECOMMENDATIONS 2-1 Seismic Capacft1es of Low and High Frequency Components 2-1 Justification for Spectral Acceleration Reduction Procedure 2-3 Equipment Evaluation Case 2-4 Equipment Design Case 2-8 Results of Study 2-9 Reco11111endations for Reducing Ground Response Spectra 2-13 3 SIMPLIFIED ANALYSIS PROCEDURE FOR EQUIPMENT MODELS 3-1 Introduct1on 3-1 General 1-00F Model 3-2 Force Viewpoint 3-4 Displacement Viewpoint 3-5 Interpretation of Ductility Scale Factor, F~ 3-5 Sliding Model 3-6 Rocking Hodel 3-13 Asymptotic Reduced Spectral Acceleration Value 3-19 Sensitivity Analysis for Parameter Variation 3-24 4 CALIBRATION OF SIMPLIFIED MODELS 4-1 Introduction 4-1 Hodel Properties and Procedure 4-Z Sliding Model 4-2 Rocking Model 4-4 Nonlinear Computer Program 4-6

DRAfT Earthquake Ttme Histories 4-8 Results of Nonlinear Analyses 4-8 Statistical Analysis to Determine Simplified Hodel Parameters 4-16 5

SUMMARY

AND CONCLUSIONS 5-1 6 REFERENCES 6-1 APPENDICES A PROPERTIES OF EQUIPMENT ANCHORAGE A-2 Introduction A-2 Bolted Anchorage A-2 Welded Anchorage A-3 Weld Properties Used 1n Study A-7 References A-10 B DEVELOPMENT OF SIMPLIFIED DYNAMIC PROCEDURES {2-DOF) B-2 Introduction B-2 S11d1ng Model B-2 Normalized Secant Frequency Squared, X 8-6 Fraction of Mass Hissing From Dynamic Mode B-7 Spectral Acceleration Capacity B-8 Ratto of Dynamic Mode Base Force to Total Base Force B-9 Effective Damping Ratto 8-11 Damping Ratto at the Maximum Displacement B-11 Damping Ratio at Effective Frequency Ductility Scale Factor Determination of Effective Frequency and Damping Procedure for Calculating Ductility Scale Factor ExillDPle Calculation Q-17 B-17 B-19 8-19 B-20 Rocking Model B-28 Nonnal1zed Secant Frequency Squared, X B-32 Fraction of Hass Hissing From Dynamic Mode 8-33 Spectral Acceleration Capacity B-36 Ratio of Oynutc Mode Base Factor to Total Base Force B-37 Effective Damping Ratio 8-39 Ductility Scale Factor 8-40 Determination of Effective Frequency and Damping 8-41 Procedure for Calculating Ductility Scale Factor 8-42 Example Calculation B-46 References B-46 viii

C EXAMPLE BUILDING RESPONSE AMPLIFICATION D EARTHQUAKE RECORDS USED IN STUDY ix

ILLUSTRATIONS t

    .Ei9JLct                                                                      fu.l 1-1      Comparison of Example Uniform Hazard'Response Spectrum with            1-2 NUREG/CR-0098 Response Spectrum {5 Percent Damped}

1-2 2-1 2-2 Incoherence of Ground Motion Comparison of Design. Yield and Ultimate Response Spectra for Low Frequency and High Frequency Components Steps in Determining HCLPF Capacity Basis for Determining Adequacy of Reduction Procedure for 1-14

                                                                                   .2-2 2-5 2-3                                                                             2-6 Evaluation Case 2-4      Basis for Determining Adequacy of Reduction Procedure for              2-9 Design Case 2-5      Example of Reduced Response Spectra for Uniform Hazard               Z-10 Evaluation Spectrum Based on Models Using t~e Simplified Methodology 2-6     Example Reduced Response Spectra for UHS Anchore<i to 0.3 g          2-15 PGA 3-1     Pseudo-Linear Elastic Formulation for One-Degree-of-Freedom            3.3 Cantilever Model 3-Z     Model for Component Sliding                                            3-6 3-3     Derivation for Normalized Secant Frequency Squared. x. for One-    3-8 i.

Degree-of-Free<iom Sliding Hodel 3-4 Flow Chart for Calculating Response Spectrum Re<iuction Factor, 3-10 Sa/Sa 1, Using Simplified Methodology for the Sliding and Rocking Models , 3-5 Response Spectra Used tn the Example Calculation 3-15 3-6 Model for Component Rocking 3-16 3-7 Derivation for Normalized Secant Frequency Squared, X, for One- 3-17 Degree-of-Free<ioai Rocking Model 3-8 Exilll!ple Reduce<i Response Spectra for Uniform H~zard Evaluation 3-25 SpectrUII Based on Sliding Model for Different Permissible Ultimate* Weld Ojsplacement . X

3-9 ExaDTple Reduced Response Spectra for Uniform Hazard Evaluation 3-25 Spactrwa Based on Rocking Modal for Different Permissible Ultimate Weld Displacement 3-10 EXi111Ple Reduced Response Spectra for Uniform Hazard Evaluation 3-26 Spectrlllll Using Sliding Model for COllbinat1ons of Mass Ratte,** and Coefficient Friction, 3-11 Example Reduced Response Spectra for Uniform Hazard Evaluation 3-26 Spectrum Using Rocking Hodel for Different Mass Ratios, a 4-1 Design Response Spectrum for Calibration Analyses 4-3 4-2-. -~lifi~g Modal Properties 4-5 4-3 Rocking Model Properties 4;7 4-4 Response Spectra for Earthquake Time Histories Used in Study Normalized 4-9 to O.Bg Peak 5-Percent Damped Spectral Accelerat1on A-1 ExaJIIJ)le Force-Displacement Curve for 3/4 in. Wedge Type Anchor Bolt A-4 in Tension, f'cu3700 ps1 {Concrete Spall1ng Failure Mode) A-Z Example Force-Displacement Curve for 3/4 in. Wedge Type Anchor Bolt A-4 in Shear, f'c*3700 psi (Shear Failure Through Threads) A-3 Fillet Weld Schematic A-5 A-4 Normalized Weld Curves A-7 A-5 Equivalent Elasto-Plastic Function for 3-16-Inch Fillet Weld I-Inch A-9 Long Loaded Transversely Using E60XX Electrodes A-6 Equivalent Elasto-Plast1c Function for 3-16-Inch Fillet Weld 1-Inch A-9 Long Loaded Longitudinally Using E60XX Electrodes B-1 Hodel for Collll)onent Sliding 8-3 B-2 Sliding Hodel Modes at Ult1mata Response B-4 8-3 Force-Displacement Diagram for Weld B-5 B-4 Sliding Hodel Oyna11ic Mode Forces Adjusted to Effective Frequency B-8 B-5 One-Quarter Cycle of Displacement in Weld at Ultimate Response B-14 8-6 Hysteretic Loop in Sliding Hodel Weld B-15 B-7 Flow Chart for Calculating Response Spectr11111 Reduction Factor, B-21 Sa,./Sa 9 , Using Simplified Methodology for Sliding and Rocking Models B-8 Response Spectra Used 1n the Exlllll)le Calculations B-22 B-9 Hodel for Coraponent Rocking B-28 B-10 Rocking Modes at Ultimate Response B-29 xi

I

• B-11 Dynamic Mode Free Body Diagrams for Determining Fraction of Missing B-34 Hass, F14 B-12 Rocking Model Dynamic Mode Forces Adjusted to Effective Frequency B-36 C-1 Shear Beam Building Hodel C-3 C-2 Target and Artificial Time History Ground Response Spectra C-4 C-3 Comparison of Response Spectra at the Top Mass and C-4 Ground Levels 0-1 Scaled R.G. 1.60 (Artificial) Earthquake Time History and Response 0-4 Spectra 0-2 Scaled Olympia. WA Earthquake Time History and Response Spectra D-5 D-3 Scaled Parkfield, CA Earthquake Time History and Response Spectra D-6 D-4 Scaled Tabas. Iran Earthquake Time History' and Response Spectra 0-7 D-5 Scaled Imperial Valley, CA {E~C. Array No. 5) Earthquake Time o.:s History and Response Spectra D-6 Scaled Nahanni, Canada Earthquake Time History and Response D-9
• Spectra D-7 Scaled Saguenay, Canada {Site 20) Earthquake Time History and 0-10 Response Spectra 0-8 Scaled Guli. USSR (Karakyr Po~nt - East) Earthquake Time History D-11 and Response Spectra D-9 Scaled Bear Valley, CA Earthquak~ Time History and Response D-12 Spectra D-10 Scaled Gazli, USSR (Karakyr Point - North) Earthquake Time History 0-13 and Response Spectra 0-11 Scaled Saguenay, Canada (Site 16) Earthquake Time History and D-14 Response Spectra D-12' Scaled Leroy Modified Earthquake Time History and Response Spectra 0-15 D-13 Scaled Leroy, Ohio (Perry NPP Basemat) Earthquake Time History and D-16 Response Spectra 0*14 Scaled New Brunswick. Canada (Mitchell Lake) Earthquake Time History D-17

         -and Response Spectra 0-15   Scaled Artificial Earthquake Time History a~d Response Spectra       0-18 xii

DRAFT TABLES

 ~                                                                       f.l!lll 3-1  Simplified Methodology Steps for Sliding Hodel for Case of No Hass 3-11 at Base and Zero Coefficient of Sliding Friction 3-2  Example Sliding Model Analysis                                     3-14 3-3  Simplified Methodology Steps for Rocking Model for Case of No Mais 3        at Base 3-4  Example Rocking Model Analysis                                      3-22 4-1  Earthquake Records Used in High-Frequency Study                     4-10 4-2  Ductility Scale Factors, F~, for Yield Capacity Based on            4-11 Nonlinear Time History Sli ing Madel 4-3  Ductility Scale Factors, F~, for Ultimate Capacity Based an         4-12 Nonlinear Time History Sliijing Model 4-4  Ductility Scale Factors, F~, for Yield Capacity Based on            4-13 Nonlinear Time History Roe ing Model 4-5  Ductility Scale Factors, Ft, far Ultimate Capacity Based an         4-14 Nonlinear Time History Rae ing Hodel 4-6  Ratios of Ductility Scale Factors (Nonlinear Time History Analysis  4-17 to Simplified Procedure) and Statistics for Sliding Hodel Yield Cap1c1ty 4-7  Ratios of Ductility Scale Factors (Nonlinear Time History Analysis  4-18 to Simplified Procedure) and Statistics for Sliding Hodel Ultimate Capacity 4-8  Ratios of Ductility Scale Factors (Nonlinear Time History Analysis  4-19 to Simplified Procedure) and Statistics for Rocking Model Yield Capacity 4-9  Ratios of Ductility Scale Factors (Nonlinear Time Htstory Analysis  4-20 to Simplified Procedure) and Statistics for Rocking Model Ultimate Capacity 4-10 Ductility Scale Factors, F,, for Scaled Nahanni Earthquake Designs  4-23 and Nahanni Input Based on Nonlinear Time Htstory Sltdtng Hodel 4-11 Ductility Scale Factors, F,, for Scaled Nahanni Earthquake Designs  4-24 and Nahann1 Input Based on Nonlinear Time History Rocking Hodel xiii

4-12 Ratios of Ductility Sc~le Factors (Nonlinear Time History Analysis '4'-25 . to Simplified Procedure} and Statistics for Scaled Nahanni Earthquake Design for Sliding Hod~l 4-13 Ratios of Ductility Scale Factors (Nonlinear Time History Analysis 4-26 to Simplified Procedure{ and Statistics for Scaled Nahanni Earthquake Design for Rocking Mode 8-1 Cutoff Frequencies Used in Study B-12' ' 8-2 Simplified Methodology Steps for Sliding Model B-23 8-3 Determination of Ratio of Total Base Force to Dynamic Hoda Base Force, B-26 R. at Frequency, f, and Damping Ratio, ~ (Used to Obtain Ry and Ry) B-4 B-5 B-6 D-1 Example Sliding Hodel Analysis Simplified Hethodologf for Rocking Model Example Rocking Hodel Analysis Earthquake Records Used in High-Frequency Study B-27 B-43 B-47 D-3 t ' xiv

DRAFT Section 1 INTRODUCTION OBJECTIVE OF STUDY This report presents the results of a proJect conducted for the Electric Power Research Institute (EPRI) to investigate the potential for damage from high-frequency seismic ground motions on nuclear power plant {NPP) components. The study was performed to develop a rational procedure for reduc1ng input to high-frequency components that are being evaluated for earthquake response spectra rich tn high-frequency energy. The final results from the study are intended for use in the Individual Plant Examination for External Events {IPEEE) to resolve NRC Severe Acc1dent Policy (SAP) issues (1). In addition, the results will be useful for reducing site-specific design response spectra with significant high-frequency energy. BACKGROUND ON HIGH-FREQUENCY SEISMIC MOTIONS The results of recent seismic hazard analyses conducted for the eastern *u.S.(EUS) indicate that ground motion response spectra have relatively large high-frequency spectral acceleration content above approximately 10 Hz. This is in contrast to typical NPP design or seismic margins-type input, such as the NRC Regulatory Guide (R.G.) 1.60 (Z) or NUREG/CR-0098 {1) ground response spectra. However, these same EUS ground 110tt*ons have significantly lower spectral acceleration content at the low-frequency end of the response spectrum (i.e., between about 1 and 10 Hz), which indicates to the seismic engineering community that these motions are less damaging than traditional design or evaluation-type input. Figure 1-1 shows the relationship betwe~ the median NUREG/CR-0098 response spectrum anchored to 0.3g peak ground acceleration (pga) and an example unifol"'III hazard spectra (UHS), with a 10,000 year return period (at the a.as fract1le), from I hazard analysts using the EPRI methodology for a EUS nuclear power plant site. Uniforra hazard for the EUS spectra are in sharp contrast to typical plant design or evaluation spectra that have been used in the past by engineers to assess the seismic safety of nuclear power plants. Design response spectra have not had significant energy in the high frequencies. These design spectra have their peak spectral accelerations at frequencies less than 10 Hz and typically return to the 1-1 V

011.00 0 (/) z 0 NUREG/CR-0098

                     ~

• Example EUS UHS et:

                    *w                                                       I

_J wo.so u I u N <( _J

                     <(

et: l-u w o_ (/) 0.00 1 10 100 FREQUENCY, hz Figure 1-1. C011p1r1son of example un1foni haz1rd response spectrum with NUREG/CR-0098 (l) response spectrua (5 percent damped).

DRJJF1 peak ground acceleration at frequencies between 25 and 35 Hz. The shapes of these design spectra are based upon the characteristics of western United States (WUS) earthquake ground 1110tions that have been of sufficient amplttude and energy to be potentially damaging to engineered structures and/or equipment. In regard to high-frequency seismic input, the following two questions must be addressed.

1. What are the potential seismic safety consequences to NPPs fl"OIII EUS ground motion response spectra that substantially exceed the plant's design spectrum at frequencies above about, 10 Hz?

2. How should EUS UHS or si,te-specific response spectra be modified in order to produce response spectra to be used by engineers in their evaluation or design of NPP structures and equipment?

The second question is the subject of this report. With regard to the first question, two points must be emphasized. First, the authors know of no case in which a well-engineered, re,inforced concrete or steel structure has ever been significantly distressed by earthquake ground ll'IOtion, except when the 5 percent damped spectral accelerations exceeded 0.20g at frequencies well below 10 Hz or the spectral velocity exceeded 5 inches/second. This statement is applicable even when such structures had little or no seismic design. Furthermore, significant distress refers to significant visible cracking of concrete or yielding of steel and is well short of failure. Dainge correlates best with spectral accelerations below 4 Hz. These statements are also applicable to anchored industrial (co111111ercial) grade equipment typ1cal of that used in power plants, even when neither the anchorage nor the equipment were designed for seismic-induced forces. Consc1entious searches for exceptions to these statements have been conducted by several investigators with no success to date. Correlation of damage to industrial facilities and engineered structures with various ground motion parllll8ters are reported in References 4 and 5. In fact, Reference 4 defines potentially damaging ground motion as being ground motion that produces 5 percent damped spectral accelerations in excess of 0.20g at frequencies less than 10 Hz. In addition, the ground motion must have adequate duration. Second, the empirical evidence--that damage to a well-engineered structures or anchored equipment will occur only if the ground motion produces substantial spectral accelerations at frequencies below 10 Hz--is supported by analytical studies. For a concrete or steel structure with a lateral load-carrying system (shear walls, braced frame, etc.) to be severely distressed, it must undergo sign1ficant interstory drifts in at least one story. In 110st cases, interstory 1-3

~rifts must exceed 0.4 percent of the story height before one might reasonably expect severe distress. Thus. for a IO-foot story height, interstory drift must exceed about 0.5 inch *. As the structure drifts and bec01118s nonlinear, its effective frequency 1 , is lowered: Any structure undergoing at least 0.4 percent tnterstory drift will have an effective frequency of less than 5 Hz by the time such drift is reached. Unless the input ground snotion'produces sufficient spectral acceleration below 5 Hz to overcome the spect~al acceleration capacity of the structure, it will be illll)ossible for the structure to develop at least 0.4 percent interstory drif~. The.issue of the effective frequency of structures being lowered as the structure drifts nonlinear has been extensively studied, with the resul~s presented in References 6 and 7. For example in Reference 7, a.r,ecent rigorous study of the Diablo Canyon Turbine Building was, conducted for the Oiablo Canyon seismic prpbabilistic risk assessment (PRA). In this study two hundred nonlinear time-history analyses were performed on a dynamic model of the Diablo Canyon Turbine Building *using 25 different ground motion time histories, with each being scaled to 3 'different amplitudes. This building had a fundamental frequency of about 8 hz. One purpose of this study was to define the characteristics of the ground motion that were required for this gro~nd motion to be poten~ially duaging to the turbine building. It was found that to be potentially damaging (interstory drift in excess of 0.4 percent), the ground motion had to produce 5 percent spectral accelerations that exceeded all three of the following spectral acceleration, s., limits: Hiqb-freauencv Lfm1t s.> 1.6g with1n 8.6-to 9.5 Hz range M1d:freguency Lim1t*

         . S,> 3.2g within 2.8 to 6.0 Hz range Low-freauencv Lfmft s.> 2.75g within  2.4 to 3.5 Hz range Unless the high-frequency limit was exceeded at some frequency in the 8.6 to 9.5 Hz range, the structure reaained elastic and did not begin to soften due to 1

nonlinear drift. Unless the mid-frequency limit was exceeded, nonlinear drUts stopped at very low interstory drift levels. Last, unless the low-frequency limit 1The concept of effective frequency is discussed in Section 3 of this report. 1-4

_DRAFT was exceeded within the 2.4 to 3.5 Hz range, interstory drifts of 0.4 percent could not be reached. The conclusion is that there fflUst be substantial spectral acceleration content over a broad frequency range and that this frequency range must extend to frequenc1es below 3.5 Hz for the ground 1110tion to be potentially damaging to this structure. Most nuclear power plant structures have fundamental frequencies in the 1 to 10 Hz range and have sufficient interstory drift capacity to drop their effective frequency below 4 Hz before severe distress will develop. Based upon nonlinear studies such as presented in References 6 and 7 and other similar studies, it is w1dely accepted among knowledgeable eng1neers that ground motion must have substantial spectral acceleration content in the 1 to 4 Hz range to be potentially damaging to nuclear power plant structures, and that higher spectral accelerations at frequencies in excess of about 10 Hz are of little or no interest in the design of such structures, unless a very severely poorly designed and constructed link exists within the structure. It has also been observed that damage correlates well with elastic spectral acceleration in the 1 to 4 Hz range. Thus, it is likely that it is an "effective* frequency in this range as opposed to the fundamental frequency that is important to s1gn1ficant damage. These results support the conclusion that it is the low-frequency content of earthquake records which causes d~ge. One example of a high-frequency earthquake that occurred near a nuclear facility was the Leroy earthquake (magnitude Hi_ 5.0) which took place January 31, 1986 about 11 miles from the Perry Nuclear Power Plant (!). At the time of the event Cleveland Electric Illuminating Company was within days of core load and receipt of their 5 percent power license for Perry. The staff observed no indications of damage to systems in operation at the time, which was confirmed by follow-up inspections. However, after the earthquake the analysis of the 1110tion records indicated that the Operating Basis Earthquake (OBE) ground response spectru~ had been exceeded 1t frequencies above 10 Hz, and the Safe Shutdown Earthquake (SSE) response spectrWI had been exceeded above 15 Hz. Tha maximum acceleration of the ground motion at the foundation level was 0.18g with a corresponding peak 5 percent damped spectral acceleration of about O.Bg; however, the peak ground velocity was only 0.9 inch per second, and the maximum ground displacement was only 0.06 inch. The 0.9 inch per second velocity and the 0.06 inch displacement indicate low energy content in the 1 to 4 Hz frequency range; thus, it is not surprising that no damage occurred. In addition, the cumulative absolute velocity 1-5

(CAY) value for this event. which is a measure of dura,tion. was only 0.08g*sec which is much lower than the OBE exceedance criterion limit of 0.30g*sec {!). In the analysis and design of nuclear power plant structures it is widely felt that to be damaging. earthquakes must be r1ch in spectral content 1n the 1 to 4 Hz frequency range. Because most nuclear power plant structures and equipment have fundamental frequencies significantly above 1 Hz. strong amplified motions below 1 Hz caused by soft soil conditions, although potentially dill'llaging to flexible structures such as high-rise buildings, do not affect typical nuclear power plant sites. For stiff nuclear power plant structures at typical soil or rock sites it is necessary to have energy content present in the l to 4 Hz frequency range for earthquakes to be potentially damaging. It 1s believed that the damaging frequency range for equipment probably is slightly higher than the 1-4 Hz range for structures. but below 10 Hz. As shown tn Figure 1-1 the NUREG/CR-0098 spectral shape has more damage-potential characteristics than does the exal'llple UHS for either equipment or structures. Experience with high-frequency motions from other environments indicate that they are much less damaging than their low-frequency counterparts. In the development of a criterion for determining the exceedance of the OBE. experience concerning the damaging characteristics of high-frequency motions was collected and documented(!). Based on a review of ground motions caused by conventional high explosive blasts, and their effect on structures. it was concluded that no damage will occur to engineered structures and equipment for short-duration ground motions with spectral accelerations below the envelope of the OSE response spectrum and a threshold cracking spectrw1 (established from blast data). However, it was not established in Reference 4 whether these conclusions can be extrapolated to longer duration large-magnitude high-fn1quency events. The conclusions that damage is not caused by high-frequency 1DOtions is also supported by fragility data for ductile equipment, perfon1ance of equipment under industrial environment vibrations and code requirements for high-frequency loading. Earthquake experience demonstrates that well-secured high*frequency c0111ponents perfon11 well during seisic events. Thus the conclusions in this report are supported by empirical evidence as well as the theoretical studies presented herein. Host equtp111ent within a nuclear power plant have fundamental frequencies in the S to 25 Hz range. In so111e cases equipment fundamental frequencies substantially in

DRAFT excess of 25 Hz are calculated. However, this situation occurs only when the analyst has either ignored or mistakenly estimated the flexibility of the attachment of the equtl)lllent to the structure. It ts nearly impossible to mount a piece of equipment sufficiently rigidly in a structure to produce a funaamental frequency substantially in excess of ZS Hz at high shaking levels, unless the equipment has trivial mass and thus is not vulnerable to seismic-induced distress {relay chatter may be an exception). Thus, for equipment, the primary concern ts spectral accelerations up to about 25 Hz. All but the most brittle equipment has sufficient distortion capability to quickly drop its effective frequency to well below 25 Hz before 1t is severely damaged. In most cases, the attachment of the equipment to the structure must undergo a minimum distortion of at least 0.10 inch before its load transfer capability is fully mobilized. A simplistic but informative way to look at the damage capability of input motion to equiplll8nt follows. No element with)n the load transfer path will undergo an intra-element distortion in excess of the spectral displacement content of the input motion at the effective frequency of the equipment when it undergoes this distortion. AssU111ing ill input spectral acceleration of l.Og, the spectral displacements corresponding to various frequencies are: SoectraJ D1soJacements corresponding Frequency to 1,oq Spectral Acceleration 5 Hz 0.39 inch 10 Hz 0.10 inch 25 Hz O.OZ inch For most practical cases, even ff all of the distortion ts concentrated into a single element, this elu,ent will not fail ff its distortion capabi~1ty exceeds the spectral displacement of the input 1110tion at the effective frequency of the equipment. Since the weak-link elements of most items of rugged industrial equipment can acc0111110date distortions of at least 0.1 inch, spectral accelerations of less than lg at frequencies of 10 Hz and greater are unlikely to result 1n equipment failure. Therefore, it is reasonable to conclude that the damaging frequency range for most items of equipment is less than 10 Hz, with the exception of relay chatter during shaking. Based on past experience, both from earthquakes and other physical pheno111ena that 1-7

produce high-frequency motions, it is believed that high-frequency input, which is a predominant part of EUS response spectra such as shown 1n Figure i-1, will not be damaging to ducti*le structures and equipment. This report addresses any situation that can accomaiodate at least 0.01 inch inelastic distortion before failure occurs. There is still a potential concern that these 1110tions may be an issue for acceleration sensitive equipment such as relays, contactors, ~otor starters and switches. For this category of components the effects of high-frequency motions should be addressed only if it is determined from a systems perspective that functional failure (e.g., relay chatter) during strong 1110tion is detrimental or equally that the safety function affected can not be confidently recovered after an earthquake. This report does not address the issue of relay chatter. The methodology in this report is based on a hypothetical conservative bounding case where a very strong piece of equipment is attached to the supporting structure through a stiff but weak anchorage. In this cas~ the entire load path except for the anchorage remains elastic, and all of the nonlinear distortion must be accolllllOdated within the weak anchorage. Based upon a review of common anchorage schemas, it was concluded that fillet welds loaded transverse to the welds represented the anchorage approach with the least distortion capability (see Appendix A). The lfm1t1ng 0.01 inch inelastic distortion assumed in this report accoanodates the smallest practical weld as well as bounding other components that are installed 1n NPPs. In general th1~ is a very conservative case for.study. A larger more realistic non-recoverable distortion would produce even larger reductions. OVERVIEW OF HIGH-FREQUENCY COMPONENTS AND BASIS FOR STUDY Host high-frequency components in nuclear power plants are very strong. If c0111ponents have high fundamental frequencies (i.e., relatively stiff c0111pared to their mass) they also have high strength, sfnce stiffness and strength are correlated. In addition to being strong most high-frequency components in nuclear power plants also have ample ductile capacity to resist the small displacements associated with high-frequency motions. Examples include the casing on a pump 1110tor, piping and the exterior shell on a small heat exchanger. These types of COIIPOnents are also ductile, and small displacements associated with high-frequency motions can be safely accoamodated if the yield capacity is exceeded. Also, ,any seismic anchor motions that correspond to the lower structure frequencies (i.e., less than 10 Hz) are not an issue since they are analyzed as 1-8

part of the normal design process and are not affected by high-frequency spectral content. A class of high-frequency components that have potentially limiting capacities are stiff electrical cabinets supported at their base, where the anchorage system is potentially the weak link (e.g., welded or secured with expansion anchors). An investigation was performed to find exa111ples of real electrical cabinets that have high fundamental frequencies based on actual testing (i.e., 15 Hz and higher). However, the high-frequency cabinet examples that were found had very strong stiff supports (overstrength welds or many anchor bolts). In this sludy a limiting case is postulated where an electrical cabinet is assumed to have a high fundamental fjxe<l-base frequency, and the anchorage is sized exactly to resist the 1nput earthquake. The anchorage strength was no larger than required to resist the design earthquake. From this viewpoint limiting components can be thought of as having high fixed-base frequencies (i.e., 10 Hz and higher) with min1mu~ strength anchorage. This class of components was assumed to represent the critical case for all high-frequency nuclear power plant components, other than relays, and was investigated 1n detail. In general, the models analyzed in the high-frequency study are very conservative relative to the general population of high-frequency components found in nuclear power plants. This perspective should be kept 1n mind in reading this report. The only source of nonlinear behavior that was assumed to occur in the study was the anchorage system at the cabinet base. Other factors that would also reduce the load on the anchorage system include the following:

• Nonlinear response in the cabinet (e.g., slight buckling in the side panels or local nexibility in the cabinet material near the anchorage)
• Nonlinear response at the connection between the electrical devices and cabinet shell
• Nonlinear response tn the devices themselves
• Reduction to high-frequency motions in nuclear power plant structures caused by incoherence of the ground motion (see discussion below)

These sources of reduction exist for real electrical cabinets; however, 1t was conservatively assU111ed in the study that all potential nonlinear behavior is concentrated in the equipment anchorage. This assumption allowed the study to 1-9

                                                                              - _..... _.. J systemat1~illlY investigate the var~ou,s parameters that influence the res*ponse to high*frequency motions. The disadvantage of this ilpproach 1s that the additional m~rgin that exists for high.frequency components was not quantified.

It was determined early in ,the investigations that welded anchorages are more brittle than bolted*anchorages. Expansion anchors have ultimate displacement capactttes in the range of approx1mately 0.1 to 0.5 inch, while small fillet welds loaded in shear were found to have corresponding capacities in'the range of about 0.01 to 0.1 inch. It 1s clear fr011 the results of this study, and past tnvesttgations (~), that seismic capacity ts limited by the displacement capacity .of,the weakest element. -Thus, all the analyses were performed assum1ng welded

'anchorages. Using a small welded anchorage represents a major squrce of conservatism in the study.

As discussed tr Appendix A and *based on Reference 8, the non*.recoverable distortion capability of a fillet weld loaded transverse to the, weld ts about 5.6 percent of the weld size. Qther investigators have reported somewhat greater distortion capilb111ty, but the capacity from Reference 8 was used in the study. Th* smallest weld that would be used to.anchor a structural base of a piece of equipment would be 3/16 inch. Such a weld would have a *mfnilllUID distortion capability of about 0.01 inch. Somewhat smaller welds might be used to anchor a 'sheet metal cabinet, but in this case, the sheet metal would u~dergo more than 0.01 inch distortion prior to weld failure. It 1s concluded that the least

• credible non*recoverable distortion capability of any anchorage is 0.01 inch.

The effect of nonlinear anchorage distortion is to lower the effective frequency of the equipD19nt and to increase the energy dissipation capability (2). Both effects are c011110nly defined 1n terms of a ductility scale factor (also called inelastic energy absorption factor), F,, by whic~ the elastic computed deaand (spectral acceleration demand) can exceed the yield cap1city*(spectr1l acceleration capacity) without failure. Then, to 111&fntafn a consistent factor of safety, the elastic spectral accelerations for an evaluation (or design) earthquake input Sa 1 associated with a fixed-base frequency f, and damping~, should bf! reduced by a ductility scale factor F,, to obtain a reduced spectral acceleration Sar. Thus: ' (l*l) 1-10

 ~ote that F~ is dependent on f, and~,. as well as on other parameters as discussed in this report.

For a 5 Hz piece of equtpme~t subjected to ground motion. the lowering of the effective frequency and increase in energy dissipation dua to a 0.01 inch non-recoverable anchorage distortion is negligible. because the elastic distortion of the 5 Hz piece of equipment ts many times greater than 0.01 inch. Thus. at 5 Hz, F is approximately equal to 1.0 and there ts no benefit from inelastic anchorage distortion. Therefore, engineers have come to treat welded anchorages as "brittle" and have traditiona.lly defined design spectral acceleration as being equal to the elastic spectral acceleration. However, for a 25 Hz piece of equipment, the lowering of the effective frequency and increase of energy dissipation due to a 0.01.inch nonlinear anchorage distortion ts substantial because such a nonlinear distortion js no longer s~all when c~rnpared to the elastic distortion of the equipment. For EUS response spectra such as in Figure 1-1, both the lowering of the effective frequency and the increase of energy dissipation associated with a 0.01 inch nonlinear distortion are important. The net effect ts that F, is substantially greater than unity, and the evaluation spectral acceleration at 25 Hz should be much less than the elastic spectral acceleration at 25 Hz. However, for WUS ground motions that produce spectra that are similar in shape to the NUREG/CR-0098 spectrum in Figure

• 1-1, the lowering of the effective frequency below 25 Hz actually increases the spectral acceleration input, which counters the benefit gained from the increased energy dissipation, andrF, is again close to unity.

Because, in the past, NPP design response spectra have always been low-frequency content spectra, and because small nonlinear distortions such as 0.01 inch do not produce IF, significantly greater than unity for such spectra even at 25 Hz, engineers have typically not IIOdiffed such spectra. However, when dealing with a high-frequency content EUS input, the response spectrum should be modified by the appropriate F, associated with at least some minimal nonlinear distortion such as 0.01 inch. Making such a 1110dif1cat1on will produce damage-consistent spectra more si~ilar in shape to the design spectra that have been used fn the past. The basis for this reduction is discussed in detail in Section Z. SPATIAL VARIATION ANO INCOHERENCE OF GROUND MOTION The ground 1110tion response spectrum estimated from seismological studies represents the elastic response of simp~e oscillators mounted on a small

                          *o 1-11

lightweight pad ~n the free ground surface. These are not the motions "seen* by~* large massive structure founded on a substantial size baseut. *coaparison of response spectra obtained from motions.measured on t~e baseraat of large structures* with those obtained froa 110t1ons meas~red on.small lightweight pads show substantial differences. particularly at frequencies in excess of 5 *Hz (for instance, see Reference 9). At frequencies of 10 Hz and greater the spectral accelerations obtained on the basamat of large structures are o(ten more than a factor of two less than those obtained on a small pad at the free ground surface. These reductions in the spectral accelerations at higher frequencies 'input tci

• large structures appear to be due to two causes:,

1. Vertical spatial v'arat1on of the ground motion, wave scat~ering effects, and soil-structure interaction.,

2. Horizontal spatial variation.and incoherence of the ground motion.

On soil sites, the first ca4se is likely to be the greatest source of these h1gh-frequency reductions. However, on a stiff rock sae; this cause is not likely tcr result in 111.1ch reduction (probably less than a 10 percent redu~tion, unless the~ structure is deeply imbedded). Furthermore, wher-e appropriate, the reduction due 1 to first cause has generally been taken into account in the seismic: design o~ evaluation of nuclear power plants. Horizontal spat.1al var1ati~~ and incoherence of the ground motion will reduce the input at high*er frequencies to structures w1th large plan dimensions. This reduction occurs on rock sites as welJ as soil sites. It has seldom been considered in the seismic design or evaluation of nuclear power pi'ants because the collection of qbservat1ona1

• data co'ncerning 'this reduction is of recent vintage'.

Based on studi~s conducted for the U.S. Nuclear Regulatory Commission, Reference

                 '                                                    {         ,  I

• 10 presents soae reconaendlt1ons concerning the reductions of h1g~er frequency spectral acceleration input to structur_es to account for their large plan '

dimensions. These reconmendat1ons are based on a conservative interpretation.of a; small data set, of observed variations of ground motions recorded over short

• separation distances. These rec011111&ndations have been adopted for use 1n seismic margin reviews of existing nu~lear 'power plants 1n the U.S. (.ll), Guidelines recoanended by ttte U.S. Department of Energy (DOE) for use 'in the seismic design
• and evaluation of DOE facilities also provide these recomandations (lZ.), which state:

           "Horizontal spatial variations in ground ~otfon result from 1-12

r. J

nonvertically propagating shear waves and from incoherence of the input motion (i.e., refractions and reflections as earthquake waves pass through the underlying heterogeneous geologic media). The following reduction factors may be conservatively used to account for the statistical incoherence of the input wave for a 150-foot plan dimension of the structure foundation: fundamental frequency of the s011-structure system {Hz} Reduction Factor 5 1.0 10 0.9 25 o.s For structures with different plan dimensions, a linear reduction proportional to the plan dimension should be used: for example, 0.95 at 10 Hz for a 75-foot dimension and 0.8 at 10 Hz for a 300-foot dimens1on (based on 1.0 reduction factor at 0-foot plan dimension). These reductions are acceptable for rock sites as well as soil sites. The above reduction factors assume,a rigid base slab. Unless a severely atypical condition is identified, a rigid base slab condition nay be assumed to exist for all structures for purposes of c0111puting this reduction.* Since the publication of Reference 10 (1986), considerable additional observational data has become available on the incoherence of high-frequency ground ll!Otion over short horizontal plan dimensions. This data has been collected primarily from research sponsored by EPRI using a closely spaced instrument array in Lotung, Taiwan (ll). Based on this research, Dr. Norm Abrahamson has rec1ntly provided an estimate of the coherence of the ground motion at short horizontal separations for a Western U.S. nuclear power plant rock site (see Figure 1-2). From Figure 1-2 one can note that the incoherence of the ground motion at frequencies above about 12 Hz is substantial, even at separation distances as short as 66 feet. Based on this estimate of the incoherence, an analysis has been conducted using a dynamic model of an auxiliary building to compare floor spectra obtained assuming incoherent versus coherent motion. The results of this unpublished analysis suggest that the incoherence reduction factors presented-above may be slightly 110re conservative than necessary. At this time, the incoherence reduction factors presented above should b1 used. However, ongoing research may result in greater reductions being suggested 1n the future, particularly for structures with plan dimensions in the 30 to 150-foot range. 1-13

HORIZONTAL- 20 m 1 I I I ____ L _____ L _____ I ----- 1 I I I 0' 0.8 -----r-- 1

                            --r-----~-----~-----

I ,I I !Ill 0.6

          -----r----                -----r-----r-----

_____ L _____ t.: ____ L _____ I - - - - -

i: 1 I I I 8 -----~-----~--

1 I

                                         --~-----~-----

I I ~ 0.4 ----r-----r---- r-----r----- g ------------------ 1 1 I I I -----~----I I u 0.2 -----~-----~-----~--- 1 I I

                                                        -~-----
          -----r-----r-----r-----r-0 0           s           10           lS           20             25 FREQUENCY (Hz)

HORIZONTAL- 50 m 1------------------------ I I I I

 -              --r-----r-----r-----r-----
~

8 B 0.8 ffi

..J 0.6
          -----~---
          -----~---- I r - - - - - ~- - - - -
          -----~- ---~-----~-----~-----
                               -}------~--,---: - - - - -

I

                                    -----~-----~-----

I 1- - I

~    M    -----r-----r -* --r-----r-----

0 _____ L _____ I - - __ -1 _* ___ I - - - - - t:, I I I I ~ ~ -----~-----~---- 1 I I

                                                -----L----- I
          -----r-----r-----r-- --r-----

00 S 10 1S 20 25 FREQUENCY (Hz) Figure 1-2. Incoherence of Ground Motion (13.) 1-14

DMR These high-frequency incoherence reduction factors should be applied in addition to the inelastic-energy-absorption reduction factors described in this report. In fact, the incoherence reduction factor should ba applied first, since tt represents a reduction 1n the motion input to the structure and to equipment mounted therein. Then, the inelastic-energy-absorption correction as discussed tn the following sections should be applied to this reduced motion. ORGANIZATION OF REPORT Section 2 discusses the'background on the seismic capacities of low- and high-frequency equipment. This provides the motivation for reducing an evaluation or design response spectrum in the high-frequency region. Justification for the spectral acceleration reduction procedure is then discussed for both the evaluation and design cases. The final results of the high-frequency study are then provided along with recoanendations for reducing ground response spectra to be used for practical cases. Section 3 lists the equations and their bases for calculating the reduction factors that correspond to the recOD111endations made in Section 2. Tht analytical bases are given for both the sliding and rocking models that were investigated in the study. An explanation for a limit on the recommended spectral reduction is given, and results from sensitivity investigations for various model parameters are shown. Section 4 addresses the calibration of the sliding and rocking models. The models, time histories and results of nonlinear analyses are presented. The findings from the statistical analysis of the nonlinear time history and the simplified procedures are given, that comprise the basis for the two empirical frequency and damping model parameters used to calculate the high-frequency spectral reduction. Section S gives a s1J111111.ry and conclusions of the study, and references are given in Section 6. Appendices A through D provide support for the main sections of the report. Appendix A discusses the properties of bolted anchors and welded anchorages. The basis for the properties for welded anchorage used in the study is provided. Appendix B gives the detailed derivations of the sliding and rocking models that were calibrated from the results of the nonlinear time history analyses. Note 1-li

that Section 3 gives the modified equations to be used in practical applications where UHS rich in high-frequency energy should be reduced. Appendix B supports Section 3. Appendix, C provides exilllP_lt analysis results for amplification of building response. This supports the required ~tnimum safety factors that are required for the reco11'1118nded spectral reduction procedure. Finally, Appendix O shows the time histories and corresponding response s~ectra for the earthquake records used tn the study. Currently a test program sponsored by EPRI is underway to demonstrate the physical behavior of sliding and rocking models secured by small fillet welds when subjected to earthquake motions. The nonlinear time-history analysis computer program that was used to develop the simplified approach in this report is also being benchmarked against the test results. 1-16 J

• DRAFT Section 2 BASIS, FINDINGS AND RECOfit1ENDATIONS This section gives the basis that motivates the procedure for reducing response spectra in the high frequency region. The physical behavior of both low* and high-frequency equipment is discussed, and justification for the reduction factors is prov1ded for both the evaluation mode (e.g., when perfonn,ng a Seismic Margin Assessment) and the design mode (i.e., when a component is being designed for a specified input). A summary of the results is given without detail (see later sections for the technical development) and is followed by recommendations for implementing the reduction procedure for practical cases encountered in review or design of components in NPPs.

SEISMIC CAPACITIES OF LOW ANO HIGH FREQUENCY COMPONENTS It is the prevailing belief in the se1sm1c engineering co11111Unity that welded anchorages are brittle. The earthquake experience that supports thts conclusion is associated with low-frequency components damaged by low-frequency seismic motions. The damagab11ity of low frequency earthquakes is also supported by recent analytical studies (.6., I). Figure 2-la shows an example case for a "low.frequency" component (f.e., frequency f 1 ) which is welded at its base. The lower response spectrum in this figure represents the design input to the component (i.e., the input used to size the welded anchorage) and the upper response spectrWI is the scaled input corresponding to .lz21h the yield and ultimate capacities. Two important points are demonstrated by Figure 2*la for low-frequency components with brittle anchorages:

1. Once the seismic input ts scaled to the level corresponding to the yield capacity (t.e., the limit of elastic response) the ultimate capacity is for practical purposes at the same level.

2. The factor of safety for the component ts equal to the ratio of the mld_ capacity to the design capacity.

Item 1 ts another way of saying that the component ts brittle. By definition brittle means that when the component reaches the limit of elastic capacity (1.e., yield limit} there is little or no additional capacity beyond that level (i.e., 2*1

0 U) 0.1 1 10 FREQUENCY, f(hz)

a. Low-Frequency Component 0

U) 1 10 1* 100 FREQUENCY,f(hz)

b. High-Frequency Co!llponant Figure 2-1 C011parison of Design, V1eld and Ulti~ate Response Spectra for low-Frequency and High Frequency Co!llponents 2-2

DRAFT the ductility scale factor is unity). Item 2 implies that the factor of safety for fillet welds (i.e., in the range of about 2) was established by code colllllittees with a conscious realization that additional capacity beyond yield can not be relied upon. However, based on failure tests welds do have some ductile capacity. Although the displacement limit for welds is small it has a very important effect in increasing the capacity of high-frequency components. Figure 2-lb shows an example case for a "high-frequency* component (i.e., frequency fh) which is welded at its base. Again, the lower response spectrum represents the design input, but here there are two higher response spectra, that correspond to the yield and ultimate capacity levels. Two important points are demonstrated by Figure 2-lb for high-frequency components with "brittle" anchorages:

1. The ultimate capac1ty for a high-frequency component can be significantly higher than the yield capacity.

2. The total factor of safety for a high-frequency component consists of two parts:

• A safety factor equal to the ratio of the yield capacity to the design capacity, that is similar to the factor provided for low-frequency components.
• An additional factor, equal to the ratio of the ultimate capacity to the yield capacity (i.e., ductility scale factor),

that does not exist, or is very small, for low-frequency components.

• For high-frequency components this additional ductility scale factor is a *bonus*

beyond the margin provided by code allowable weld stresses. Nonlinear analyses can not practically be perfon,ed for every component as part of the standard design process. Thus, code COl!IDittees have established factors of safety that reflect the design practice. Recognizing the brittle behavior for welds used to anchor low-frequency components, code writers have provided allowable weld stresses anticipating that failure will occur at essentially the yield level. As demonstrated in Figure 2-lb this is conservative for high-frequency components. The presence of this extra ductility capacity provides a rational basis for reducing an evaluation (or design) response spectrum for high-frequency components. JUSTIFICATION FOR SPECTRAL ACCELERATION REDUCTION PROCEDURE There are two perspectives from which response spectrum reduction can be viewed. 2-3

In the evaluation mode, such as a Seismic Margin Assessment (SMA) review, a capacity 1s calculated and generally expressed in terms of a ground response spectrum parameter [e.g., peak ground acceleration (PGA) or a response spectral ordinate at a specified frequency]. For example, in SHA high confidence of low probability of failure (HCLPF) capacities are calculated based on an input ground response spectrum shape such as a NUREG/CR-0098 response spectral shape anchored to a PGA. When a HCLPF is reported for a component 1n terms of the PGA it 1s implied that the entire response spe~trum shape is scaled along w1th the PGA. In the evaluation mode the ground response spectrum is scaled up or down to define the individual equ1pment capacities. In contrast, in the design mode the ground input response spectrum is fixed and all equ1pment and structures are designed to that input. In des1gn the ground response spectrum is stationary and the supporting elements are sized to that input. Justification for the response spectrum reduction factors can be made in terms of either the evaluation or the design modes. The same procedure for reducing ground response spectra is found from e1ther perspective. The individual minimum safety factors are in general different becausi of differing ph1losoph1es associate with each mode. In the following two sub-sections justification for the reduction procedure is given for the evaluation and the design cases. Eaujpment EyaJuatjon case The ductility scale factor provides a rational basis for reducing an evaluation response spectrum that contains significant high-frequency energy. In order to motivate the discussion that follows the steps to determine a HCLPF capacity spectrum are first shown in Figure 2-2. This process starts with Step l where the spectral capacity of a component mounted on the ground is calculated. Then in Step 2 the Sar capacity is represented by a point on a response spectrum plot. Step 3 then shows the HCLPF capacity response spectrum, that is anchored to the Sar capacity point. This would be the case when no reduction is considered. Finally in Step 4 the Sa capacity is anchored to the reduced portion of the HCLPF capacity spectrum (it is assumed that it is already known). This effectively J:.l1lll the HCLPF capacity of the component since the HCLPF capacity response spectrum ,snow higher. The simple process in Figure 2-2 shows the benefit of taking credit for the extra capacity due to ductility for a high-frequency component. 2-4

DRAFT SiGp..l Dctcn:iuce Sa capKUy, Sa, Fy ll-- Sa. i::: ~ Stqi.l Pio< Sa capacny Oil & response spectrum Sa

                                       - - - - -1 Sa,
                      ' - - - - - - -f- - -... f SU::r;t.l Anchor HCI.PF shape to capacity Sa HCU'F capaaty rCllpOmNI spec:ltllm (Wllhoo.tl roduCIIOO)

I

                      .___.;.._ _ _ _ _ _f,,__ _ _....,.. £ s.tqu. Anchor~ shape to capaaty Sa                            Ha.J'Fc;apac,cy rapom,,

lpOCltllm (COlllldcnq ndltcboa) I

                      ' - - - - - - - - - - - ' : - - - - -...... f f

Figure 2*2. Steps in determining HCLPF capacity. 2-5

a-(./) z lklllld~ Ylollli.- 0

                   ~

a:: w

                   .....J wO.!ID uu
                    ~
                    .....J 1-u w

1 ~ .. ~ .... a.. 2. ,YIMd&l---*k1 (./) 0.00 +---,--,....,....,,...,...,..,.,...._-,--.-,-.,..,.,.....,..---,J.-..,.....,......,.......4

o. 1 10 100 FREQUENCY, hz Figure 2-3 Basis. for Determining Adequacy of Reduction Procedure for Evaluation Case Figure 2-3 shows schematically* how th1s concept can now be applied to reduce an evaluation response spectrum (e.g. an EUS UHS defined to be a SMA seismic margin

.-earthquake). An example component with a frequency f is visualized which has a spectral acceleration capacity, Sa,, that is plotted as a point on Figure 2-3. Its capacity is evaluated following the rules that lead to a HCLPF (11). For example, if the component is a squat item of equipment located on the ground that is controlled by anchorage capacity and is subjected to pure base shear, then Sa/g is just equal to f/Wt. H,re g is the acceleration due to gravity, FY is the SHA yield capacity and Wt is the weight of the component. Note that the yield capacity, Fy, would be detennined based on a SMA procedure such as Reference 11: For example, if welded anchorage is used, which is a brittle failure mode for low frequency components, then the conservative deterministic failure margin (CO~) capacit1 is defined at about the 99 percent probability level (ll). This implies that there is a factor of safety, F111 , between the calculated SHA yield capacity and the median yield capacity (i.e., at the 50 percent probability level) of about 1.5. Also shown on Figure 2-3 is the HCLPF capacity spectrum and its associated reduced response spectrum, that has been scaled 'to coincide at frequency f with spectral capac1ty Sar. Note that it is the*reduced portion of the spectrUIII that 1s 2-6

superimposed on Sa,. In this demonstration it is assumed at this point in the discussion that the reduced response spectrum has already been obtained and is being checked for its adequacy. Note that the reduced response spectrum merges with the capacity spectrUTD it low frequencies where no reduction is justifierl. A higher response spectrum is ilso shown in Figure 2-2 that is the HCLPF response capacity spectrum (i.e., the HCLPF capacity spectrum without any reduction) scaled to the median yield level as determined based on Sa 1

• For lower frequencies this upper response spectrum also corresponds to the ultimate capacity as discussed above (see Figure 2-la). But at higher frequencies the reduced spectral value can be used to anchor the HCLPF capacity response spectrum, as long as the ultimate capacity corresponding to the reduced spectral value is equal to the median yield level obtained from the non-reduced portion of the HCLPF capacity spectrum.

As shown in Figure 2-3 the ultimate Sa capacity comes from the reduced spectral ordinate, Sa,, and includes both the safety factor, FSM, between the reduced spectrum and the median yield level as well as the factor due to the additional ductile capacity {i.e., ductility scale factor, F~)- In contrast, the median yield capacity shown is based on the non-reduced capacity spectral ordinate, Sa 1 , and includes only the SMA capacity-to-median yield safety factor. This is the total safety factor at low frequencies, where there is no reduction. At low frequencies the total safety factor is due entirely to the ratio of the median yield capacity to the SMA capacity {i.e., the ductility scale factor is unity) consistent with the SMA philosophy. In equation form the reduction factor, Sa,.1Sa 1, can be obtained by equating the ultimate Sa to the yield Sa as discussed above: (2-1) From which the reduction factor is found to be: Sa,./Sa 1

• 1/F, (2-Z)

Equation 2-2 says that the reduction factor is just inversely related to the ductility scale factor. It is important to note that F, is a function of the median yield capacity of the weld based on the reduced spectral input, that is 2-7

DRAFT equal to Sar FSII. Since Sar is not know a priori, an iterative process is required in general to calculate F~- The example demonstration in Figure 2-3 verifies that the reduction for a component with a frequency, f, is adequate. In general, the development of the reduced portion of the response spectrura requires that F, be calculated at various frequenc1es in order to develop the entire reduced curve. The reduced curve which is produced is conservative for any SMA HCLPF that is detennined to be less than the seismic margin earthquake (SME) input, since F, increases for capacities below the SME. However, it would be non-conservative for capac1 t 1es above the SME, or for i n-str,ucture response spectra that are greater than the ground-level input response spectrum. As discussed in Appendix C for structures the spectral amplification between the ground and up in a structure above about 10 Hz is less than 2. Th'us, 'the minimum factor of safety used for developing F, should be increased by a factor of 2.0 when performing a SMA as discussed below (see Section entitled RESULTS OF STUDY). Eau1pment Des19n Case Figure 2-4 demonstrates schuatically how the reduction scale factor concept can be applied to reduce a design response spectrum (e.g., a design earthquake response spectru11 with high frequency energy content). An example design response spectrum is shown in Figure 2-4, below which a reduced spectrum is constructed. It is assumed here that the reduced response spec~rum already has been determined and its adequacy is being checked. The reduced spectrum merges with the design spectrum at low frequencies where no reduction is justified. The adequacy of the reduced response spectrum is demonstrated schematically for a component with a frequency, f. The top response spectrl.llll shown in Figure 2-4 is the design spectrum (without any reduction) ,scaled to the median yield level based on Sa,,. For components with lower frequencies this upper response spectrum also corresponds to the ultimate level as discussed above (see Figur~ 2-la). But at hig~er frequencies a reduced spectral value (1.e., Sar instead of Sig) can be used to design components as long as the ultimate cap1city corresponding to the reduced spectral value 1s equal to the yield level obt1ined from the design spectrum. Note in Figure 2-4 that the ultimate Sa capacity shown is based on the reduced spectral ordinate and includes both the safety factor between the reduced spectrum and yield as well as the fact9r due to th~ ad~itional ductile capacity (i.e., ductility scale factor). In contrast, the median yield capacity shown is based on 2-8

DRAFT 0"11.oo 0 (/) z-0

                 ~

Cl'.: w

                 ..J wo.so u

u<(

                  ..J
                  <(

Cl'.: I- uor.-aa- .. a.., u 1 w 2. Y l e l d l l a - .. ll&o Q._ (/) 0.00 o1 1 10 100 FREQUENCY, hz Figure 2-4 Basis for Determining Adequacy of Reduction Procedure for Design,Case the original design spectral ordinate, Sii,, and includes only the design-to-median yield safety factor. This is appropriate since at low frequencies, where there is no reduction, the total safety factor is due entirely to the ratio of the yield capacity to the design capacity {i.e., the ductility scale factor is unity) as intended by the coda writers. Similar to the arguments given in the previous subsection the same reduction factor is obtained {see Eq. 2-2). RESULTS OF STUDY As described in the following sections nonlinear time history analyses of models that represent cabinets anchored with welds were conducted. The purpose of these analyses was to determine the additional inelastic ~nergy absorption factor and to calibrate the parameters of two simplified analytical methodologies developed to predict the response of equipment that could slide or rock. These methodologies can be used to determine the reduction factors to apply to an evaluation {or design) response spectrum without having to perform detailed nonlinear time history analyses. Figure 2-5 shows example reduced response spectra for an example site UHS. The reduced spectra shown were developed using the simplified methodologies for the 2-9

                                                           ' 4 '

s '----o.eo c:3 f i Cf.I ~AnchcnQe Olllam**

• 0.01 lnc:h SllllaGMmllll M)'fflllCIICIGRaduoadog,a .*

                                                                               ~Mlldli

Ji* G.OII "-o T *0.S I h/w - 5 FSM

• 1.5 0.1 ~ 10 20 50 100 Frequency, f, , (Hz)

Figure 2-5

• Example of Reduced Response Spectra for Uniform Hazard Evaluation Spectrum Based on Model, Using Simplified Methodology sliding and rocking models that are discussed in Section 3. These models represent cases where the failure mode of a component is either sliding or rocking of the oase that causes the connecting weld to fail. The purpose here is to concisely indicate the final results without the b~rden of explaining' the details A of the 111&thodologfes., This explanation is presented in Section 3 (and Appendix W B).

Two evaluation response spectra are shown in Figure 2-5: one anchored to 0.3 g pga and a.second spectral shape anchored to 0.6 g pga., Below each evaluation spectM are reduced spectra based on the simplified methodologies for sliding and rocking. In general, it was found that the sliding model reductions are less .than the rocking model reductions. However, it is possible that the rocking case may control for some shapes of ground response* spectra, part.icularly at higher frequencies and high spectral input (see top case in Figure 2-5).' Notice that the reduced spectra extend below the pga, and eventually they will become asymptotic to a* "reduced pga" at ave~~ ~igh frequency' (off the scale of Figure 2-5). It should be noted that the frequency plotted along the horizontal ax1s corresponds to the frequency of the equipment ib.!2ll the anchorage (i.e., fixed L 2-10

                                                                         .. - )

base). As the equ1pment frequency becomes higher and higher the total system frequency (i.e., equipment plus anchorage) eventually reaches a finite value. Th* system frequency approaches the case of a rigid mass (i.e., the equipment) supported on a spring (i.e., the anchorage). Because of the anchorage stiffness the reduced response spectrU111 can be below the pga of the unreduced response spectrum at high frequencies. Three general recommendations are made using the results from the studies performed to generate Figure 2-5. The bas1s for these recorrrnendations is discussed below.

• When developing time histories to be used to calculate in-structure response spectra the reduction factors should be based on a response spectrum equal to a factor (that 1s equal to or greater than 1.0) times the evaluation ground response spectrum. In other words, the evaluation ground response spectrum should first be scaled by the factor, the reduced response spectrum obtained, and then the resulting reduced spectrum divided by the factor. The final reduced ground response spectrum would then be used to develop in-structure spectra.

The factor to be used depends on whether a design or evaluation is being performed and whether the in-structure response spectra is at the ground level (or below) or up in the building. Recommended factors ara g1ven below.

• The calculation to obtain the reduced 9round response spectrum should be stopped at a minimum frequency of 10 Hz and the reduced spectra reconnected to the evaluatton (or design) response spectrum at 8 Hz.
• An asymptotic reduced spectral accelerat1on value can be utilized at high frequencies so that the reduced response spectrum can be based on only the simplified methodology for the sliding model'. This will avo1d the problem of the reduced spectra, based on the two models, crossing as shown at the top of Figure 2-5.

From Figure 2-5 it is observed that the a1110unt of reduction for the UHS response spectrum anchored to 0.6 g is less than the reduction for the spectrum anchored to 0.3 g. This occurs because the frequency shift caused by nonlinear response assoc1ated with a given weld distortion (e.g., 0.01 inch) is less at higher capacity levels. This implies that as ~n evaluation response spectrum is scaled up, the reduction factors approach unity. The fundamental natural frequencies of real nuclear power plant buildings are generally less than 10 Hz; hence, the reduction process will effect the response 2-11

of the second and higher building modes. For practical situations 1t is expected that amplified in-structute response spectra at frequencies above about 10 Hz will be less than twice the ground response spectra (see Appendix D). This is the reason that in-structure response spectra should be obtained using 1nput time histories developed from a reduced response spectrum cal~ulated based on a higher spectru~ (i.e., greater than a factor of 2.0). Stopp1ng the reduction at 10 Hz (and tapering to the evaluation spectrum at 8 Hz) will ensure that the reduction process does not propagate into the amplified spectral acceleration region near the fundilllental frequency of the butld1ng. In general, the reduced response spectrum to be used to develop 1n-structure response spectra should be based on reduction factors obtained from both the sliding and the rocking models. As a general rule for equipment mounted up in a building, an evaluation spectrum should be scaled in small increments between a factor of safety of 1 to the required minimum factor of safety, and the corresponding reduction factors detennined for both models. The envelope of the resulting reduction factors should be used to scale down the evaluation response* spectrum in the high-frequency region. Note for a seismic evaluation (i.e., SHA) a required minimum factor of 3.0 should be used that consists of the factor between the HCLPF and median yield capacity (i.e., 1.5) and the increase for higher mode building ampl1ficat1on (i.e., 2.0). For design a minimum factor of safety of 4 should be used that con$1sts of the factor between design and median yield capacity (i.e., 2.0) and the increase for higher mode building amplification (i.e., Z.O). Both the sliding and rocking 110dels also should be investigated to obtain the reduced response spectrum for use in analyzing equipment mounted at the ground. For the cas, of equipment 1110unted at the ground only the factor of 1.5 for a SHA and 2.0 for design is required (i.e., the extra factor of 2 for building amplification is not needed). In lieu of this procedure a conservative alternate, that is easy to implement, can be used. Since both 1110dels are asY111Ptotic to a connon reduced pga, that is always higher than the spectral acceleration at which the curves for the two models cross (e.g., see Figure 2-5), the minimum value can be set at the asymptotic value and only t~e sliding model used. For S percent damped ground response spectra a conservative value for the asymptotic pga 1s always equal to the peak evaluation 2-12

DRAFT spectral accelerat1on obtained at 10 percent damping divided by a factor of 1.6. It is recorranended that this value be used; although, slightly greater reductions can be obtained using the detailed incremental procedure descr1bed above. For cases where the asymptotic value is above the pga the minimum of the un-scaled spectral acceleration and asymptotic value can be used. The basis for the asymptotic limit is given in Section 3. It was found in the study that the design of welded anchorage leads to safety factors between the design and yield levels that vary depend1ng on parameters such as the design level, component mass distribution, coefficient of friction and model type (i.e., sliding or rocking). From the results of the study as demonstrated in Figures 2-3 and 2-4 as discussed above, higher factors of safety always lead to lower spectral reduction factors. It was found for sliding lllOdels that the minimum factor of safety for ruj_gn is about 2.0, while the safety factor for rocking can be as low as 1.5 for practical cases where a component only rocks. In a SMA the capacity of a component is generally evaluated at a conservative yield level. For th1s case the safety factor is about 1.5. As discussed above components up in a building should be evaluated for an in-structure response spectrum based on reduction factors obtained from a ground response spectrum scaled by an additional factor of 2. This is equivalent to doubling the factor of safety. In sunvnary, this implies for a SHA evaluation of existing equipment that a required safety factor, F514 , of 1.5 should be used for components mounted at the ground and a value of 3.0 for components up in a building. Similarly, the minim1.111 factor for the sliding case in the design mode is 2.0, and the reduction factors should be based on a F511 of 2.0 for components mounted at the ground and a value of 4.0 for components up in a building. In order to s1mplify the analysis a value of F111 equal to 4 can always be conservatively used. The use of this value has only a relatively moderate effect on the resulting reduced response spectra. RECOMMENDATIONS FOR REDUCING GROUND RESPONSE SPECTRA It is reco11111ended that spectral reduction factors used to reduce an evaluation or a design response spectrum for analyzing high-frequency components be obtained using the simplified sliding model presented in the next section. As discussed below it is recommended that the fraction of mass at the base and the coefficient of friction both be set to zero. The following guidance 1s given when applying 2-13

these recol!Dllendations: I. The reduction should be performed for frequencies above 10 Hz. and the reduced response spectrum should be reconnected to the evaluation spectrum at 8 Hz.

2. The 5 percent damped reduced response spectrum should not be reduced below a response spectrum value equal to the peak evaluation spectral acceleration at 10 percent damping divided by 1.6, or the elastic spectral acceleration at frequencies greater than that at which the elastic spectrum peaks, whichever is less, unless additional cases are considered as discussed above.

3. In the evaluation mode for equipment mounted at the ground the minimum factor of safety, Fvi, should be equal to 1.5, and for equipment up 1n a building rSM should be 3.0.

In the design mode for equipment mounted at the ground Fgi should be equal to 2.0 and for equipment up 1n a building FSM should be 4.0. Note that FSM equal to 4.0 can always be conservatively used for all cases. A permissible anchorage distortion of 0.01 inch should be used to develop the reduction factors. This corresponds to the ulti~ate non-recoverable displacement capacity of a 3/16-inch weld. It 1s assumed in selecting a 3/16-inch weld that there possibly may be 1/8-inch welds use<I to attach the sides of electrical cabinets to embedded floor plates. However, there are other sources of flexibility in this type of component that are equivalent to the models used in the high-frequency study which assumed 3/16-inch welds. Thus, the use of a 3/16-inch weld size also represents these cases. The authors believe that a pemissible non-recoverable anchorage distortion of 0.01 inch is conservative and that all nuclear power plant c0111ponents have at least this minimum amount of displacement capacity. For plants that can justify greater permissible anchorage distortions, it may be possible to base the spectrum reduction factors on these larger permissible anchorage distortions. However, the capacity of electrical cabinets anchored with nom1nal welds in the ~lant as well as other components must be considered in justifying larger permissible distortions. The analyst should realize when performing a SMA evaluation using reduced response spectra as input that the porti-0n of the reduced response spectrum above 8 Hz takes partial credit for ductile capacity (0.01 inch nonlinear distortion). In perform1ng a SMA the inelastic energy absorption factor recommended in Reference 2-14

11 also may be used, in general, for high frequency components since these factors are based on the characteristics of WUS ground motions. Note that Reference 11 recommends F~

• 1.0 for welds and other small distortion capability anchorages so no additional credit beyond a reduced response spectrum is taken for brittle components.

Figure 2-6 shows example reduced ground response spectra for.,,, a UHS anchored to 0.3 g pga based on the sliding model following the above recorrmendations. The spectra based on F511 values of 1.5 and 3.0 would be used for an SHA evaluation for equipment at the ground and for developing in-structure response spectra, respectively. The spectra in Figure 2-6 based on FSM of 2.0 and 4.0 are examples of reduced spectra for use in design where the required minimum design to yield is determined to be 2.0. The reduced response spectrum corresponding to FSM equal to 2.0 would be used for design of equipment mounted at the ground and the top one (i.e., F511 equal to 4.0) would be used for developing in-structure design response spectra. Note that the curve based on F511 equal to 4.0 could be used for all cases since it is conservative and only moderately different from the other three curves. - §

           ~

1 UHS Evalulllon I Specilnn1

                                                                  ~---ug ca l      P*mlallle ~

Olapa,oern<<lt*O.01

                  ~  *0.05 0.1 10         20       50        100 Frequency, f, (Hz)

Figure 2-6 ' Example Reduced Response Spectra for UHS Anchored to 0.3 g PGA. 2-15

The shape of the reduced response spectra in Figure 2-6 are very similar to the type of ground response spectra used in the past to design and evaluate nuclear power plants (e.g., R.6. 1.60 and NUREG/CR-0098 response spectral shapes). Thus, the UHS ground response spectra currently being obtained for Eastern U.S. sites have similar characteristics to traditional design spectra when the UHS ara modified*to have consistent safety 111argins across all frequencies. Finally, the reduction factors can be developed "si1111larly for both horizontal and vertical ground response spectra. Although the behavior of a component that fails due to vertical motion will involve a racheting response (i.e., larger downward nonlinear displacements compared to upward), experience shows that t~ese types of components in NPPs have additional capacity available for earthquake loading compared to vert i ca 1 components subjected to hori zonta 1 1cads. Fir.st, these components are always found to be very conservatively anchored to walls (or devices in cabinets anchored ~o panels). Second, if.they were designed.to just barely meet allowable stresses, there would be extra capacity available for earthquakes because of the margin inherent against dead load, that can be used. 2-16

Section 4 CALiBRATION OF MODELS INTRODUCTION The procedures for us1ng the sliding and rocking models are presented in Section

3. In developing these models two empirical parameters were defined and calibrated from the results of detailed nonlinear time history analysis. The frequency parll'll8ter "a* determines the rel at i onsh_i p between the normaJ i zed effective frequency squared, x., and the normalized secant frequency squared, X, that is expressed in the following fonn for both the sliding and rocking 1110dels:

(4-1) The second parameter is the damping parameter "b" that is used to estimate the hysteretic damping value evaluated at the secant frequency based on a full force/displacement loop that assumes that both the positive and negative displacements are equal to the ultimate displacement. This is physically unrealistic since it is unlikely that a full hysteretic loop would be effective at ultimate response. The relationship for the effective damping ratio that includes the parameter bis as follows: (4-2) 1.- where:

                            ~h

• Maximum hysteretic damping evaluated at the ultimate displacement

                            ~,

• C0111ponent elastic damping ratio at fixed-base frequency The approach for calibrating parllll8ters a and b was to perform nonlinear tima history analyses and to utch the results using the simplified methodologies by varying the parameters a and b. This was dona systematically, and separately, for the two models, and values of a equal to 1.6 and b equal to 0.3 were found to be 0

4-1

DRAfT

  ' "best fit* ,results for both the sliding and rocking models; The model properties and procedure used are discussed in the next sub-section. This is followed by a discussion of the *earthquake time histories used in the analysts. Finally, the results of the statistical analysis of the simplified procedure and the nonlinear analysts results are given and compared to previous results from other invest ig.ations.

HODEL PROPERTIES ANO PROCEDURE The general 2-DOF models for, component sliding and component rocking are shown in Figures 3-2 and 3-6, respectively. Each of these models includes a portion of the mass at the base to include the effects of the rigid-body mode. Both the simplified procedures and the models for the nonl.inear analysts are based on these e* models. As discussed in Section 2 it ts recCl1!JIHnded that ,the mass at the base of the 1110dels be set to zero. However, in the calibration *nalyses masses were applied at the base and a realistic coefficient of friction was provided in the sliding model in order to determine values of the a and b parameters which fit real situations. The following is a description of the models and the procedure 1 used to select th e properties' for the sliding and rocking models. This is followe<.I by a description of the nonlinear computer program used to perform the analyses:

• SUding Model Figure 4*1 sh~s the response spectrum used to design the ~elds for 6 sliding models with the fixed-base frequencies as shown, that range between 2 and 25 Hz.

Four percent damping was assumed as the percentage of component elastic damping at '1 the fixed-base frequency. The response spectrum shown. is a R.G. 1.60 spectral I shape anchored.to 0.1 g pga. A mass was assumed for the sliding model. and the weld for each of the six IIOdels was designed using a 3/16-inch transverse weld constructed with E60XX electrodes. The allowable design capacity per inch was asswned to be 3.B kips (i.e., 3/16 inch

• 0.707
• 1.6
• 0.3
• 60 ksi) *. The WiJ:.e. mass was multiplied by the spectral acceleration tn sizing the weld, which is consistent with standard design practice. No credit was taken for friction in design. Note that the absolute size of the total mass used in the calibratfon analyses is n~t significant since the results are invariant 1n terms of component size.  :,

In the*calfbration anal~ses for the sliding model 20 percent of the mass was 4-2

CJ) 1 a (/)

       ...                     4'KDamplng z

0

     ~

n:: w _J .,.. w 0.1 ~ Sa, (g) u 2 .290 u<( I w 5 .317 8 .297 13 .215 18 .164 _J 25 .128

    <(

n:: I-u w o._ (I) 0.01 0.1 1 10 100 FREQUENCY, hz Figure 4-1. Design Response Spectr1a1 for Calibration Analyses

placed at the model base. This assumption implies that the component will respond as a shear beam. In the nonlinear time history analyses a friction element was placed at the model base. This element attaches the model to the ground until the friction force (i.e., Wt) is reached. When the friction force begins to be exceeded the base slides until the friction force is reducea as the earthquake input changes and the base re-attaches to the ground. A friction coefficient of 0.4 was used for all cases considered. The weld spring is a elastic-perfectly plastic element. The yield capacity of the weld is 7.5 kips per inch, which is approximately a factor of 2 higher than the design capacity (see Appendix A). At the yield force the displacement in the weld is 0.001 inch, and the ultimate displacement 1s 0.0105 inch (see Appendix A). The properties of the sliding models are shown in Figure 4-2. Rocking Model The component was assumed to have an aspect ratio of 5 to 1. This ensures that the component will rock without sliding for a realistic coefficient of friction at the base. The design input for the six rocking cases 1s the same as for the sliding cases {see Figure 4-1). A vertical acceleration equal to 2/3 the peak ground acceleration (i.e., 0.067 g pga) was also assumed. This implies that the model is rigid in the vertical direction. A unit mass and unit height for the rocking component were used. Again, the absolute size of the mass, and the model height are not significant since the results are invariant in ter111S of the size of these parameters. The design force in the weld was determined following a standard design procedure. The vertical force in the weld was reduced for the effect of gravity. The entire mass was placed at mid-height of the component (i.e., center of gravity) and was multiplied by the spectral acceleration. The effects of the horizontal and vertical earthquake responses were combined by the 100-40 rule (1). In the nonlinear analyses for the rocking model 25 percent of the mass was placed at the model base. This assUlll)tion corresponds to a straight line fundamental mode shape that is in between the lllOde shape for a cantilever beam and a pure shear beam, which is realistic for this type of component. In order to preserve* the overturning 110111ent about the base the top russ was positioned at a height equal to 2/3 the height of the component. Note that*an incremental mass IIIOlll8nt of inertia was added in order to preserve the total mass moment of inertia of the 4-4

M: TotalConlpoMnl Kr:=:!s=- 1(_.:W.id~

                                                      .: P'radlon ...... BaN
                                                      +:  Coofldenl ol Slldlng Fl1dloll
                                                    ,, : Component Eladc
                                                          ~Rllllloal Rwld-8aN ~

di II : ~ Dua ID Gl1MIJ .....I

                                            ~

U'I F1xed Base Weld Yield Frequency (1-*) Hg

• Hg Displacement (Hz) lKtp) illJtl. ...t_ Jr- (K1~inl (K~ltnl (fn) 2 0.8 0.2 0.4 0.04 0.327 566.6 0.001 5 2.043 619.3 8 5.231 580.3 13 13.813 420. I 18 26.482 320.4 25 0.8 0.2 0.4 0.04 '51.085 246.2 0.001 figure 4-2. Sliding Model Properties

component about its corner. At each side of the 1110del base two elements 1n parallel were provided. The weld spring element shown in Figure 4-3 had the same unit properties as used for the weld spring in the sliding model. The second element is a gap element that prevents the base from penetrating across the plane of the ground. This constrained the model to rock about its two corners. The properties of the rocki_ng model are shown in Figure 4-3. Nonlinear Computer Program The DRAIN2D computer program was used to perform the nonlinear time history analyses (ll), A special gap-friction element was developed by Professor Graham H.. Powell for the project and implemented into the program. Tangent-.stiffness proportional elastic damping was asswzied. This means that during the time period th~t the' weld yields the elastic damping in the weld is zero. In general, stiffness-dependent damping was assumed for elastic damping. However, for some low-frequency cases. and particularly for earthquake time hf stories with low spectral input relative to the peak ground acceleration, the rigid-body mode was severely over damped with stiffness dependent damping. For these cases both stiffness- and mass-dependent damping was required. A time step equal to 0.0005 seconds was used for most of the analyses. The analyses results were confirmed using smaller time steps for selected cases. One practical problem' was encountered in performing the DRAIN2D analyses. When running the ORAIN2D program a scaled earthquake time history is input to the model and the displacements in the weld are calculated. The objective of the analysis is to find the earthquake scale factor that causes the weld ~o displace exactly an amount equal to the defined capacity limit (i.e., yield level at 0.001 inch or ultimte level at 0.0105 inch for a transverse weld). This was necessary since the scale factor versus weld displaclilllent relationship tends to be very flat at the yield level and above. As discussed 1n Appendix A, the yield level was determined from the actual force-deflection diagram for the weld and changad for each maximum displacement. To be consistent with the assl1lll8d yield level' the maximum displacement was fixed (i.e., 0.001 inch or 0.0105 fnch). If randOII displacements (corresponding to constant scale factors) are sought, they will generally be very different from each other. Thus, calculating target displacement required an iterative procedure. Since the relationship between 4-6

M : Tallll ~ Mas.a Kt : Coalpona,tflud.BaM~ Kw: w.ldSIIIIINa

                                                                  *=     FracllonMaNalBaN y :   Nodll HeighllColnpone Height 4Mc,: 11111 .. wur Mau Moment of 1ner1a llr:  Colnpanef1I Elallc Damping Raio
                                                                        .SfludBue~

D : Ai:.AMllllun Due lu Oravlly Hllle:Baedalalla-upandad lurdarlly; helgbl ..

                                   --->--...l-------L--L---*zag

b. Rodm1g Model I

Fixed Base Weld Yield Frequency (l-1)Hg *"9 4"o w yh Displaceaent (Hz} {Ktpl {Ktpl (Ktp-tn-sec2l --tr-- Uol Unl (ki~hnl {ki~tnl {tnl 2 0.75 0.25 0.08618 0.04 20 67 0.307 524.2 0.001 5 1.916 651.2 e 4.904 556.8 13 12.950 223.l 18 24.827 160.2 25 0.75 0.25 0.08618 0.04 20 67 47.892 123.l 0.001 figure 4-3. Rocking Hodel Properties

displacement in the nonlinear range and earthquake scale factor are not always* monotonically related obtaining ~onvergence was I practical problem. Brent's method, which involves root bracketing, bisection and Jnverse quadratic interpolation, was used (ll). By starting with two earthquake scale factors that bracketed the scale factor corresponding to the desired weld displacement, convergence 1s guaranteed by Brent's method. For each case considered ft normally took about 5 to 10 iterations to reach the desired displacement. EARTHQUAKE TIME HISTORIES

     ~ifteen earthquake time histories were.used in t~e calibration analyses. Table 4-1 lists the time histories that were used and general information for each record.

The plots of the time histories and the corresponding response spectra are given

  *, in Appendix 0. Note that Record 12 (Leroy Modified) ts the same as Record 13
   -,(Leroy, Ohio) except that the time step has been increased by the factor,, 1.25,
    'which causes the co'rrespondtng response spectrum to peak near 18 Hz (as* comp'ared
     ~o 23 Hz for Record' 13). Record 15 is an artificial ti~ history with significant ,

high-frequency content (see F1gur~ 0-15}. Figure 4-4 shows a composite of five-percent damped response s*pectra for all 15 time histories, where each response spectrum has been scaled such that its peak value ts 0.8 g. This figure indicates that significant spectral content is

• covered by t*he records between 2 and 25 Hz. High-frequency,. low-frequency and broad-banded real earthquake *time histories are included in the.set of 15 records.

By us1ng these 15 time histories and the 6 frequencies models between Zand 25 Hz, all practical cases ,are ~overed 'necessary to confidently calibrate the emp1r1ca1 a and b parameters. RESULTS OF NONLINEAR ANALYSES Tables 4-2 through 4-5 g1~,e the d'uctil 1ty scale factors, F11 , for the sl id1ng and rocking IIOdels for the yield and ultimate capacity levels based on the, ORAIN2D time history analysis. Tables 4-2 and .4.3 give F11 values for the yield and ultimate displacement levels in the weld, respectively, for the sl1d1ng 1DOdel. As explained tn Section 3 the reference yield st,te ts defined in terms of the fixed-base mo~~l and the yield

     ~ in the weld. When the force in the weld is equal to the y1eld force and the.

friction force ts equal to +Mg {just at the instant the base, begins to slide) the 4-8

1.00 , - - y - - , - , r r - , r T T r r - - r - - r - - r ~ ~ - ~ - - - - - - ~......... Ol 0.90 - z 0 0.80 - I-: -

                ~0.70 -

It.I _, 0.60 - LtJ u 0.50 - U I

                <(

,a 0.40 - _J

                <(

0:: O.JO I-f;. 0.20 (}_ (I) 0. 10 - 0.00 - -~~~~~~::_m-rrrrr--.----.-.-rm 1 0.1 1 10 100 FREQUENCY, I1z Ftgure 4-4. Response Spectra for Earthquake Time Hfstortes Used in Study Nonaal1zed to 0.8g Peak 5-Percent Damped Spectral Acceleration

Table 4-1 EARTHQUAKE RECORDS USED IN HIGH-FREQUENCY STUDY Station t,\agnitude Site Intensity NlL. Eu:tbmakt Ds1te Hill!!!! IH rii;tJ 1m Ms lf,IJ' I R.G. 1.60 (Artificial) 2 Olympia, WA 04/13/1949 Hi~hway Test Labs *N86E 7.0 VI 11 3 Parkfteld, CA 06/27/1966 Cholame No. 2 N65E 6.4 VII 4 Tab.s, Iran 09/16/1978 Tabas Trans. 1.7 X

'/

5 Iapertal Valley, CA 10/15/1979 E.C. Array No. 5 140 6.9 VII .. 6 Nahannt, Canada 12/23/1985 Site I Long . 6.9 IX 4", t> 0 7 Sagu~nay, Canada 11/25/1988 Site 20 Long. 6.0 8 Gazlt, USSR 05/17/1976 Karakyr' Point East 7.0 IX 9 Bear Valley, CA 09/04/1972 Melendy Ranch N29W 4.3 VI 10 Gazli, USSR 05/17/1976 Karakyr Point North 1.0 IX 11 Saguenay, Canada 11/25/1988 Site 16 Long. 6.0 12 Leroy Hodi fied 13 Leroy, Ohio 01/31/1986 Perry NPP Basemat South 4.8* V 14 New Brunswick, 03/31/1982 Mitchell lake 28 4.0* IV C;mada . 15 Artlffctal "s

• Surface wave magnitude
• ffo11.ent magnitude

                                                                                                                          ~fCJ
                                                                                                                            ~

Table 4-2 DUCTILITY SCALE FACTORS, Fµ, FOR YIELD CAPACITY BASED ON NONLINEAR TIHE HISTORY SLIDING MODEL ltxtel Fixed-Base F~ (Hz) _L _l._ _1_ _L J_ _L __}__ ~ _L -1.Q_ _ll_ _R _ll_ -14... _.fi.. 2 1.025 I.on 0.986 0.979 0.950 1.074 1.058 0.937 1.294 0.977 0.869 0.936 1.105 1.041 I.Oil I 5 o.~ 0.946 1.117 0.987 1.002 1.005 0.978 1.002 0.993 0.934 1.076 0.94) 1.0)6 1.048 0.996 8 0.97) 1.0)0 0.999 1.030 1.046 0.952 1.043 0.94) 1.038 0.991 0.987 l.074 1.021 1.077 1.040 13 1.004 1.004 1.007 1.000 0.968 0.990 1.131 0.992 1.039 0.988 0.961 1.117 1.094 0.931 1.020 18 1.047 0.900 0.993 1.036 0.988 1.038 1.053 0.953 0.939 0.968 0.930 0.996 1.055 1.053 0.957 25 I.OIi 0.988 l.(XX) 0.996 0.992 0.968 I.DOI 0.946 0.979 0.906 0.878 0.889 0.921 0.963 l.128 i~

Table 4-3 DUCTILITY SCALE FACTORS, Fµ* FOR ULTIMATE CAPACITY BASED ON NONLINEAR TIME HISTORY SLIDING HODEL ltldel Fixed-Base FrecJJeOC)' (Hz) _L _L _L J_ _L _L _L _.IL _L _lQ_ _lL _1£. _u_ _a_ -15.. 2 1.041 1.054 1.003 1.002 0.967 1.097 1.094 0.957 l.348 1.006 l .017 1.037 1.247 1.358 1.059 N 5 1.023 0.979 1.155 1.026 1.046 l.046 1.024 1.046 l.033 0.969 J .149 0.984 1.106 1.172 1.033 e 1.040 1.078 1.047 l.105 1.123 1.025 1.102 0.9'.JJ 1.103 1.068 1.065 1.149 1.100 1.ll6 1.120 13 1.139 1.096 1.005 l.138 1.084 1.124 1.359 1.109 1.155 1.149 l.148 i.m 1.251 1.366 1.123 18 1.238 1.094 1.049 1.273 1.179 l.185 1.265 1.252 1.179 l.323 1.147 l.490 1.418 1.341 1.218 25 1.162 1.100 l.056 1.254 1.135 l.472 l.497 l.224 1.100 l.177 1.165 l .166 l.730 1.740 1.683

                                                                                                                     ~-. ..
                                                                                                                             -I
                                                                                                                     ~ *-1,  .,,
                                                                                                                      ~
                                                                                                                       ~

yield state is asswned to be reached. On the other hand, the yield capacity is defined when the displacement in the weld reaches the displacement at the knee in the elastic perfectly plastic force deformation diagrn (i.e., 0.001 inch). This corresponds to a weld ductility value, µw, of 1.0. As seen ,n Table 4-2 the F, values for the yield case are generally close to unity. In s0111e cases they are slightly below 1.0, which is caused by the spectral acceleration at the effective frequency being significantly higher than the corresponding value at the fixed-base frequency. This relative increase is sufficiently high to'offset the reduction due to the X,/X term (see Eq. 3-7). The explanation for this result for the nonlinear time history analyses is obtained from understanding the formulation of the simplified models {see Section 3). Table 4-3 gives similar results for the ultimate capacity {i.e., when the displacement in the weld reaches 0.010S inch). The F, values for the ultimate case in Table 4-3 are always greater than the corresponding values for the yield case in Table 4-2; although, they are s0111etimes very close - particularly for low frequency models. By comparing Tables 4-2 and 4-3 it is seen that:

• For low frequency fixed-base models, and particularly for earthquakes rich in low-frequency energy (e.g., Records 1,2 and 3),

there is little or no increase in capacity above the yield level.

• For high-frequency fixed-base 110dels, and particularly for earthquakes rich in high-frequency energy (e.g., Records 13,14 and 1S), there is significant increase in capacity above the yield level.

The first point emphasizes the conclusion that for damaging low-frequency earthquakes, welds for typical components with fixed-base frequencies in the 2 to 10 Hz range respond essentially in a brittle manner. In contrast, the second point implies for earthquakes* rich in high-frequency energy that high-frequency components have significant capacity beyond the yield level. Tables 4-4 and 4-5 for the rocking model lead to the same type of results as found for the sliding model. However, there is generally a10re nonlinear increase in capacity for the rocking cases corapared to the sliding cases. This is due to the leveraging effect of the displacements in the rocking modal since a unit rocking displacement in the top rocking 1110del mass causes a reduced rigid-body rocking displacement in the weld (i.e., equal to top mass rigid-body displacement divided by the aspect ratio of the model). Note that the total displacement of the top mass consists of both rocking and flexural displacements, and the interest here is 4-1~

DRAFT the portion attributed to the rigid-body rocking contribution. In contrast, there is a one-to-one correspondence in the sliding model top and bottom mass rigid-body displacement. STATISTICAL ANALYSIS TO DETERMINE SIMPLIFIED HODEL PARAMETERS Separately for the sliding and rocking models a matrix of a and b para111eter values were selected and F~ values were calculated for the same cases analyzed by nonlinear time history analysis. Values of parameter a ranged between 1.4 and 1.7 and par-¥11eter b ranged between 0.2 and 0.5. Ratios of the ductility scale factors (i.e., nonlinear analysts to simplified analysis) were calculated and means and coefficients of variation (COVs) of the ratios were obtained. This was done separately for the yield and ultimate capacity levels. It was found that a minimum COV, w1th a mean close to unity occurred for both models and both capacity levels corresponding, to a equal to 1.6 and b equal to 0.3. Tables 4-6 through 4-9 give the ratios of F, values for the analyses considered along with the statistics for the ratios for the "opttmum* a and b values (i.e., a equal to 1.6 and b equal to 0.3). For the yield level cases the mean ratios are essentially unity, and the COVs are 0.065 and 0,095 for the sliding and rocking cases, respectively. For the ultimate capacity cases the mean ratios are also close to un1ty, and the COVs are 0.106 and 0.160, for the sliding and rocking cases, respectively, The relationship between x. and X provided by Equation 4-1 1s very similar to that presented by Kennedy(~). lwan (1§), and Sozen (11, ll) for somewhat different nonlinear problems, as can be seen fr0111 the following table. X*~ X Ea. 4-1 Kennedv Iwan Sozen 0.9 0.97 0.99 0.97 0.9 0.75 0.89 0.91 - 0.95 0.92 0.75 0.6 0.77 0.74 - 0.85 0.85 0.6 0.45 0.62 0.52 - 0.70 0.76 0.45 0.3 0.43 0.38

• 0.48 0.62 0.3 0.15 0.23 0.23 0.38 0.15 4-16

Table 4-6 RATIOS OF DUCTILITY SCALE FACTORS (NONLINEAR TIME HISTORY ANALYSIS TO SIMPLIFIED PROCEDURE) AND STATISTICS FOR SLIDING HODEL YIELD CAPACITY ftldel [arthcmlsg lk!:ord Fbm-Base freqJency Hean (!kl _i__L_L____L_ _j_ _]_ _j_ _ L -1!L -1L -1.L -1L -1L _IL .u:mL 2 1.026 1.034 0.987 0.900 0.951 1.075 1.059 0.938 1.295 0.978 0.870 0.936 1.104 1.039 1.012 1.019 (0.093) 5 0.989 0.946 1.116 0.981 0.991 1.005 0.978 1.002 0.993 0.934 l.076 0.940 1.016 l.047 0.996 1.001 I (0.047) 8 0.979 l.lX8 0.996 1.016 1.017 0.950 1.026 0.939 1.036 0.988 0.984 1.071 1.018 1.074 1.037 1.009 (0.037) 13 1.008 0.990 0.994 0.9132 0.914 0.979 1.025 0.981 1.027 0.974 0.970 1.106 1.081 0.920 1.009 0.998 (0.049) 18 1.016 0.953 0.964 1.003 0.952 1.011 0.972 0.929 0.918 0.962 0.926 0.975 1.022 1.010 0.934 0.970 (0.035) 25 0.954 0.931 0.943 0.945 0.949 0.898 0.934 0.922 0.928 0.890 0.889 0.866 0.1138 0.899 1.137 0.932 (0.065)

                                                                                                                                                   ~
            ~an     0.995   0.977   1.000   0.987   0.962   0.986   0.91J9   0.952    1.033  0.955  0.952   0.982   1.022   0.998   1.l>Zl  0.988  ~

(COV) (O.OZS) (0.037) (0.055) (0.022) (0.034) (0.055) (0.042) (0.031) (0.122) (0.035) (0.072) {0.084) (0.067) (0.066) {0.059) (0.065)

                                                                                                                                                   ~
                                                                                                                                                   ~
                                                                                                                                                   ~

Table 4-7 RATIOS Of DUCTILITY SCALE FACTORS (NONLINEAR TIME HISTORY ANALYSIS TO SIMPLIFIED PROCEDURE) AND STATISTICS FOR SLIDING MODEL ULTIMATE CAPACITY lbiel Eartb 0.999 0.994 0.971 0.964 0.991 0.990 0.997 0.995 0.976 0.978 1.037 1.030 0.992 0.992 (0.019) 5 0.985 0.930 1.028 0.988 0.995 0.982 0.997 0.973 0.997 l.020 1.079 0.953 0.971 l.006 l.(X>!J 0.994 ... (0.033) -I IO 8 0.963 0.950 0.952 0.989 0.988 0.952 0:938 0.966 0.953 1.012 0.983 0. 997 1.008 0.9'33 l.038 0.979 (0.028) 13 1.119 0.885 0.920 1.023 0.958 o.~ l.044 1.042 0.968 0.979 L0I0 0.999 l.022 0.906 l.019 0.997 (0.053) 18 1.061 0.788 0.766 1.030 0.996 1.099 l.018 1.065 1.030 l.ll3 I.OJI l.137 0.895 1.095 O.Em 1.001 (0.lll) 25 l.006 0.725 O.iu5 1.027 1.002 1.042 1.191 1.123 0.925 I.Zn l.292 1.117 l.358 I.OOi 1.126 1.065 (0.168) t:,

       ~

(WI) 1.019 0.876 0.895 1.00:I 0.985 1.004 1.030 1.027 0.978 1.066 1.062 1.030 1.048 1.029 1.012 1.005 (0.053) (0:104) (0.133) (0.018) (0.016) (0.051) (0.076) (0.055) (0.035) (0.097) (0.102) (0.068) (0.139) (0.039) (0.069) (0.095) se

                                                                                                                                               =!I

Table 4-9 RATIOS OF DUCTILITY SCALE FACTORS (NONLINEAR TIME HISTORY ANALYSIS TO SIMPLIFIED PROCEDURE) ANO-STATISTICS FOR ROCKING M<X>El ULTIMATE CAPACITY lt>del farthgjake Record Fi>m-Base fmp:!l'ICY l't!ar'I (lg) , - L _L_ _L_ _L_ -L.. _§_ _]_ ___a_ __L_ __JQ_ _u_ _JL -1L --1.L _IL 1WlL 2 0.962 o.* 0.991 0.979 0.935 *o.936 0.997 0.980 0.978 0.900 0.949 0.985 1.078 1.053 0.975 0.985 (0.037) I

         .5      O.M 0.884 0.996 0.006 0.984 0.851              o.m      0.821   0.895 0.884   1.060   0.990   0.929   0.885   0.948    0.904 I

(O.CIJ7) N 0 8 0.832 0.952 0.932 0.958 1.037 0.914 0.940 o.~ 1.008 0.983 -0.554 0.965 0.751 0.916 0.795 0.907 (0.139) 13 1.112 1.094 0.812 1.220 1,265 0.962 l.t:03 0.992 1.204 1.055 1.284 0.994 0.787 0.759 0.889 1.029 (0.16) 18 l.216 0.930 0.766 1.051 1.323 0.959 1.189 0.976 J.272 0.996 0.689 0.929 0.766 0.933 0.947 0.996 (0.181) 25 1.186 0.852 0.751 1.127 1.406 l.ll0 1.353 1.070 1.28) 0 ..921 0.582 0.994 0.874 1.029 1.131 1.044 (O.D)

        ~        1.018   0.949   0.875  1.040   1.158   0.956   1.043    0.971   1.120 0.971   0.853   0.976 . 0.864   0.929   0.947    0.978 (ClJ.1)  (0.162) (0.<m) (0.117) (0.103) (0.155) (0.082) (0.177) (0.0n) (0.130) (0.057) (0.313) (0.024) (0.132) (0.104) (0.106) (0.UiO)
                                                                                                                                                 ~
                                                                                                                                                 ~
                                                                                                                                                 ~
                                                       -                                                -                                        .-1

Table 4-4 DUCTILITY SCALE FACTORS. Fp, FOR YIELD CAPACITY BASED ON NONLINEAR TIME HISTORY ROCKING MODEL ftldel a Fixed-Base frequency (Hz) _L _L _L __!__ J_ _6_ _]_ J_ _L ---1!l _J_L --1.t. _u_ -1!.. -15... .,.. 2 0.983 0.982 1.001 0.996 0.973 0.966 0.994 0.991 0.999 0.998 0.978 0.981 1.040 1.033 0.994 I 5 0.994 0.930 1.042 0.996 1.010 0.994 1.007 0.988 1.009 1.033 1.09] 0.965 0.900 1.021 1.019 8 0.983 0.959 0.982 1.012 1.0)6 0.972 0.947 0.974 0.979 1.045 0.998 1.037 l.035 ] .027 1.073

          )3   1.182  0.968  1.006  0.900 l.CM>6 I.143  1.143   1.069  J.102 1.204  1.157 1.196  1.071 l.116  1.)46 18   1.0)6  0.953  0.929  1.)09 1.)69  0.933  1.100   J .069 1.094 l.103  l.125 1.160  1.509 1.263  1.194 C,

25 1.056 0.963 0.931 0.992 1.006 l.484 1.232 1.009 1.062 0.862 ].134 0.832 1.235 1.124 J.17J

                                                                                                                    ~

!:I

Table 4-5

                    ~CTILITY SCALE FACTORS, F~, FOR ULTIMATE CAPACITY BASED ON NONLINEAR TIME HISTORY ROCKING HODEL ltldel Fixed-Base FreqJerlcy Ui?l   _l_   ......L _L    _L    _L     _L    J_      _L     _L    -19...  -1L   _R_   _n_   .....M.. -1i.

.,.. 2 l.019 l.017 l.033 1.027 1.005 0.997 1.035 1.019 l.034 1.033 1.021 1.033 1.114 l.095 1.027 5 1.116 1.028 1.134 1.193 1.133 1.118 1.168 1.108 I.Ill 1. l!Xl 1.270 1.004 1.099 1.164 l.300 8 1.302 1.286 1.185 1.293 1.511 1.280 1.228 1.288 1.270 1.806 1.509 1.444 1.361 1.535 1.447 13 l.658 1.238 1.182 2.114 1.615 2.117 1.391 1.847 1.485 3.668 3.689 4.058 1.610 2'.045 1.945 18 l.5C8 1.173 1.105 1.370 1.650 1.056 1.889 2.098 l.626 2.307 2.441 8.033 6.284 3.861 3.105 25 1.198 1.165 1.109 1.207 1.232 2.059 2.208 1.532 1.256 l.264 1.532 3.677 9.086 5.063 3.379

                                                                 .URAFT The values given by Kennedy are a function of the duration of strong motion, with the lower values applying to long-duration motion and the higher values applying to short-duration motion. The x. values from Equation 4-1 lie close to the center of the range of values given by Kennedy for a somewhat different nonlinear problem. Also, the x. values from Equation 4-1 lie between the values given by Iwan and Sozen for somewhat different nonlinear problems. Thus, Equation 4-1 seems to be consistent with previous research study results.

The values of Pe versus X provided by Eq. 4-2, where b Ph is set equal to 0.173 (1-X) as discussed in Section 3 for 1-DOF models, are again in the same range as those suggested by Kennedy (2), !wan (ll), and Sozen (.11 and lil.) for somewhat different nonlinear problems. Assuming that Pf is equal to 5 percent, the following P. values are obtained by each of these approaches for various X values: P. (Qercent) X Ea. 4-2 Kennedv !wan Sozen 0.9 6.0 5.1 - 5.9 7.6 6.0 0.75 7.3 5.3 - 7.1 8.9 7.7 0.6 8.4 6.1 - 8.3 IO.I 9.5 0.45 9.3 7.5 - 9.6 11.3 11.6 0.3 10.4 7.9 - 11.6 13.0 14.0 0.15 10.9 7.7 - 15.2 16.2 17.3 The values g1ven by Kennedy are for in-plane nonlinear response of shear-wall and braced-frame structures that exhibit severely pinched hysteretic behavior, at least for longer strong motion durations. The lower values from Kennedy are for long-duration motions while the upper values are for short-duration motions for which this pinching is not severe. One would expect~. for nonlinear anchorage behavior to be similar to the upper values from Kennedy. The differences between the Pe values rec01111118nded herein and the upper values from Kennedy will typically have less than a 4 percent effect on computed response, and so these differences are negligible. Even the differences between the damping values recommended herein and values midway between those recommended by Iwan and Sozen will typically have less than an 10 percent effect on computed -response, and so these differences are also small. Thus the P. approach recoanended herein seems consistent with previous research study results. As an additional check on the selection of the a and b values, Record Number 6 4-21

(Nahanni, Canada Site 1) was used as design input. This record w1s scaled so that 1ts_5-percent damped peak response spectral value w1s 0.8 g and 1~6 g (two cases). For.each case 6 models corresponding to the 6 fixed-base frequencies used previously were designed per the procedure described above using the 4-percent-daJIPed scaled Nahanni response spectra. Then the earthquake were records scaled until the yield and ultimate displac&111e~ts reached their limiting values (i.e., 0.001 inch and 0.0105 inch for yield and ultimate, respectively). The ductility scale factors based on the nonlinear time history analysis are given in Tablas 4-10 and 4-11 for sliding and rocking models, respectively. Tables 4-12 and 4-13 give the ratios 9f the ductility sc~le factors for the nonlinear time history analysis to the simplified procedure for sliding and rocking, respectively. The lll8an and COV statistics are given at the bottom of each table. The values shown compare consistently with the results given in Tables 4-6 through 4-9. This demonstrates that the a and b values also work when the earthquake input is scaled to significantly higher levels. 4-22

Table 4-10 DUCTILITY SCALE FACTORS, F, FOR SCALED NAHANNI EARTHQUAKE DESIGNS AND NAHANNI INPUT BASED dN NONLINEAR TIME HISTORY SLIDING HODEL Model De~ign Sgectrum Scaled to Q,8g Sa e~ak Design spectrum Scaled to I.6g Sa eeak Fixed-Base Eregueni;v rnil U~ld Cagicih Ultimate Cagacjty Yje]d Capacity Ultimate Capai;1ty .JI, N I 2 1.078 1.097 1.068 1.095 w 5 1.008 1.035 0.992 1.029 8 0.957 l.Oll 0.952 0.998 13 0.987 1.071 0.978 1.046 18 1.063 1.165 1.068 1.150 25 1.001 1. 215 1.006 1.146

Table 4-11 DUCTILITY SCALE FACTORS, F..u! FOR SCALED NAHANNI EARTHQUAKE DESIGNS AND NAHANNI INPUT BASED un NONLINEAR TIME HISTORY ROCKING MOOEL Model De~jgn SgectrYm S~i]~~ t2 Q,S<l S~ eeak Desjgn ~Rc,truM Sci]ed t2 1,§g Si e~2k Fixed-Base Er~i:rni:nc:t Ul~ J U~]d ca1rntit:t U]tj1~t~ tggacit! Yh:l!:I Cai;rntit! UJtimgt~ tan1,itI .... 2 0.969 1.009 0.962 0.991 r.> ~ 5 0.982 1.040 0.980 1.016 8 0.960 1.098 0.955 1.029 13 l.D45 1.392 1.011 1.175 18 0.813 1.045 0.846 J.074 25 1.200 2.082 1.056 1.651

• Table 4-12 RATIOS OF DUCTILITY SCALE FACTORS (NONLINEAR TIME HISTORY ANALYSIS TO SIMPLIFIED PROCEDURE) AND STATISTICS FOR SCALED NAHANNI EARTHQUAKE DESIGN FOR SLIDING MODEL Hodel De~jgn Su~ctrym Sca]eg t9 Q,SQ Si e~1k Qesjgn SR~~tCYDI Sca]~d 12 1,§g S~ ~~ik Fixed-Bise Ere1uumtr CHzl Yh:] d Cil!il~itll U]timate Cauacitll Yi~ld ~ill!a!:;itll U]tim~te ,au1titr 2 1.078 1.089 1.069 1.090

.i:,, 5 1.008 1.021 0.991 1.021 I N U1 8 0,955 0.980 0.952 0.979 13 0.984 0.970 0.976 0.988 18 1.044 l.167 1.056 1.101 25 0.984 0.854 1.000 0.903 Mean 1.009 l.014 1.007 1.014 (COY) (0.041) (0.097) (0.042) (0.067}

Table 4-13 RATIOS OF DUCTILITY SCALE FACTORS (NONLINEAR TIKE HISTORY ANALYSIS TO SIMPLIFIED PROCEDURE) AND STATISTICS FOR SCALED NAHANNI EARTHQUAKE DESIGN FOR ROCKING MODEL Hodel De~jgo ~g~trua S~g]ed tg Q,~g Si eeat Desjgo Su1~trY1 S~a]ed to l,tig Si e~1k Fixed-Base Etl!IY!H!C! (U?;) Ijg]!;J (agacjh'. ~]tjmat~ t!l!icitl Yjg]d !:;11rn~1h'. Yltillliltl C1ga~it! 2 0.966 0.927 0.961 . 0.954

p. 5 0.978 0.944 0.978 0.970 I

N 0\ 8 0.949 0.910 0.949 0.934 13 1.008 0.932 0.995 0.898 18 0.822 1.072 0.802 1.302 25 1.079 l.044 1.005 0.825 Mean 0.967 0.971 0.948 0.980 (COV) (0.08) (0.064) (0.072) (0.154)

                                                                                                               ~

iJ

                                                                                                               ~

ilffl aail

Section 5

SUMMARY

AND CONCLUSIONS High-frequency seismic motions (i.e., greater than 10 Hz) were investigated and found to be significantly less damaging than low-frequency ground motions. It is concluded that structures and equipment at nuclear power plants have additional capacity above yield to absorb the small displacements associated with high-frequency earthquake input. The lack of ductile capacity of limiting elements (e.g., small welds) when subjected to low-frequency seismic motions provides a rational basis to reduce ground response spectra at high frequencies. By maintaining a consistent safety factor between the failure level and the design capacity across all dynamic frequencies reduction factors are obtained. The simplified procedures that were developed can be used to reduce ~arthquake ground response spectra in the high-frequency region. The different types of structures and equipment used at nuclear power plants were reviewed and it was concluded that a conservative case is a 1110del that fails at the very small displacement of 0.01 inch. This case corresponds to an electrical cabinet that is anchored at its base by a 3/16-inch fillet weld loaded in the transverse direction. For this limiting example it is also conservatively ass1J111ed that there is no nonlinear response in the cabinet, in the connection between the electrical devices and cabinet shell, or in the devices themselves. Two models that represent this conservative limiting case were studied where the inelastic -response was concentrated in the connection between the model and the base support. A sliding model that considered friction between the base and the support was analyzed. It was found that the effects of friction are not significantly different fr011 the anchorage since the anchorage yield and friction forces are interchangeable. A rocking model that included the restoring force of gravity was also analyzed. However, it was found that the sliding model is generally more conservative. This occurs because the rigid-body rotation of the rocking ~odel produces relatively less displacement 1n the anchorage compared to the sliding mode. In the sliding model the rigid-body displacement and the anchorage displacement occur one to one while for the rocking model there is a beneficial leveraging effect due to the tall aspect ratio of the cabinet. 5-1

DRAFT Nonlinear time history analyses of both models were conducted using 15 different earthquake records for six model frequencies (1.e., 2 to 25 Hz). A wide range of seismic input that represents high-frequency, low-frequency and broad-banded motions were applied to the models. The results of these analyses demonstrate that high-freque~cy 110tions are less damaging compared to low-frequency motions. The increase 1~ capacity increases at higher and hfgher frequencies. Two simplified analysis procedures were developed (one for each model) using a pseudo linear-elastic approach, that avoids having to perfonn nonlinear time history analysis. The purpose for developing these procedures was.to investigate the influence of different 1110del parameters and to provide an easy-to-usa method for reducing ground response spectra in practical applications. Two-degree of freedom models were initially investigated where a part of the mass was placed at the model base to realistically simulate dynamic behavior. Equations for these models were developed and are presented in Appendix 8. Subsequently it was found that it is conservative to u~e only 1-DOF models to develop the response spectrum reduction factors. This result is reasonable since more degrees of freedom tend to spread the displacement among the many elements reducing the displacement in the anchorage. This finding simplified the procedures ev,en further. The equations for the 1-DOF models are derived directly in Section 3 and are consistent with the general deviations in Appendix B. Finally it was found that when using the sliding model with a conservative lower bound asymptote only a single analysts for each dynaatc frequency ts required using only the sliding model to produce a conservative margin-consistent reduced ground response spectrum. The simplified procedures were calibrated with the results from the nonlinear time history analyses. It was found that the same values for the freq~ency and damping empirical parameters can be used for both the sliding and rocking models. The differences between the simplified procedures and non-linear results were found to be small (i.e., COVs of 0.11 and 0.16 for the sliding and rocking models, respectively). The resulting paraaeter values were found to corapare closely with results fn:111 past pseudo linear-elastic analyses for different IIIOdels and from different investigators. The ground response spectrum reduction factors are simply equal to 1/F,, where F, ts the inelastic energy absorption factor as given in Section 3. It is 5-Z

DRAFT rec0111118nded that F, be based on the sliding model with a peniissible anchorage, distortion of 0.01 inch. The following guidance should be followed:

1. The reduction should be performed for frequencies above 10 Hz, and the reduced response spectrum should be reconnected to the evaluation spectrum at 8 Hz.

2. The reduced response spectrum should not be reduced below a response spectrum value equal to the peak evaluation spectral acceleration at 10 percent damping dtvtded by 1.6, or the elastic spectral acceleration at frequencies greater than that at which the elastic spectrum peaks, whichever is less, unless additional cases are constdered as discussed in Section 2.

3. In the evaluation mode (i.e., SMA) for equipment mounted at the ground the minimum safety factor, FSll' should be equal to 1.5, and for equipment up in a building ~114 should be 3.0.

In the design mode for equipment mounted at the ground FSM should be equal to 2.0 and for equipment up in a building_FSM should be 4.0. - The additional factor of 2.0 for the inimum safety factor, F111 , is to account for amplification up tn the building at frequencies above about 10 Hz. This ts required because the permissible anchorage distortion of 0.01 inch ts not changed with building elevation. Thts is conservative stnce at htgher elevations the anchorage size will generally increase (i.e., greater than 3/16-tnch for fillet welds). An increase in anchorage stze leads to higher ultimate displacement capacities and hence greater spectral reductions, whtch offset the increase in motion level that is used in evaluation or design. Note that it is*always acceptable to use an F111 value of 4.0 for all cases. Thts will lead to slightly conservative results compared to the individual cases above. Another important reduction for high-frequency input that is based on ground motion tncoharence should be applied first before the inelastic energy absorption factor reduction is applied (see Section 1). The reduction for ground motion incoherence ts permitted because seismic-induced motions of a large massive structure founded on a substantial size base at are reduced from the free-field 110tions. 5-3

Section 6 PEFERENCES I. J.W. Reed, et 11. Industry Approach to Seismic severe Accident Policy Implementation. Palo Alto, California: Electric Power Research Institute, November 1991, NPw7498.

2. Design Response Spectra for Seismic Design of Nuclear Power Plants. NRC Regulatory Guide 1.60, U.S. Atomic Energy Co111Dission, Directorate of Regulatory Standards, Revsion 1, December, 1973.

3. N.M. Ne~ark and w.J. Hall. Deye]opment of cr1teria For Seismic Rey1ew of Selected Nuclear Power Plants. NUREG/CR-0098, Nuclear Regulatory Comission, May 1978.

4. J.W. Reed, et al. A Criterion for Oetenn1n1nq Exceedance of the Operating Basis Earthquake. Palo Alto, California: Electric Power Research Institute.

July 1988, NP-5930.

5. M.W. Mccann and J.W. Reed. Lower-Bound Maqnjtude for Probab1J1st1c Seismic Hazard Assessment. Palo Alto, California: Electric Power Research Institute, October 1989, NP-6496.

6. R.P. Kennedy, et al. Eng1neer1ng Characterization of Ground Motion - Task I, Effects of Characteristics of Free-field Motion on structural Response.

NUREG/CRw3805, prepared for U.S. Nuclear Regulatory Co111111ssion, Hay 1984.

7. R.P. Kennedy, D.A. Wesley, and W.H. Tong. probabilistic Evaluation of the Diablo Canyon Turbine Building Seismic Capacity Usina Nonlinear Time H1storv Analysis. Prepared for Pacific Gas & Electric Company, June 1988, Report 1643.1.

8. D.F. Lasik and J.L. Kennedy. Ultimate Strength of Eccentrically Loaded Fillet Welded Connections. Structural Engineering Report 159, University of Alberta, Canada. May 1988.

9. c.Y. Chang et al. Enqineer1na'character1zat1on of Ground Motion - Task II; Observational Dati on Spatial Variations of Earthquake Ground Motion. Vol\Jllle 3, NUREG/CRw3805, Prepared for U.S. Nuclear Regulatory Comission. Febni,ry 1986.

10. H.S. Power, C.Y. Chang, I.M. Idriss, and R.P. Kennedy. Enq1neer1nq ~

Character1zat1on of Ground Motion - summary Reoort. Volume 5, NUREG/CR-3805, Prepared for U.S. Nuclear Regulatory Commission. August, 1986.

11. A Methodology for Assessment of Nuclear Power Plant seismic Harq1n. Palo Alto, California: Electric Power Research Institute, prepared by Jack R.

Benjamin &Associates, Inc., et al., June 1991. NP-6041, Revision 1.

12. R.P. Kennedy et 11. Design and Evaluation Gu1deJ1nes for Department of Energy Eac111t1es Subjected to Natural Phenomena Hazards. UCRL-15910, prepared for U.S. Department of Energy. June 1990 6-1

DMff

13. N.A. Abrahamson and J.F. Schneider. Spatial Coherence of Strong Ground Motion for Application to Soil-Structure Interaction, RP297B-l. Palo Alto California: Electric Power Research Institute. 1990

14. A.E. Kannan and G.H. Powell. DRAIN-2D; A General Computer Program for Dynamic Analysis of Inelastic Plane structures. with user's Guide.

Distributed by NISEE/Computer Applications, EERC, University of California, Berkeley, Reports No. EERC 73-6 and EERC 73-22, April 1973, (Revised August, 1985.

15. W. H. Press, et al. Numerical Recipes. Cambridge University Press, New York,.

NY, 1986.

16. w.D. Iwan. Estimating Inelastic Response Spectra from Elastic Spectra.

Earthquake Engineering & Structural Dynamics, Vol. 8, pp. 375-399. 1980.

11. P. Gulkan and M.A. Sozen. Inelastic Responses of Reinforced Concrete structures to Earthquake Motions. ACI Journal, Vol. 71 No. 12, PP 604-610.

December 1974.

18. A. Shibata and M.A. sozen. substitute-structure Method for Seismic Design in BLC.. Journal Structural Division, ASCE, Vol. 102, STl, pp. 1-18. January 1976.

6-2

DRAFT Appendix A PROPERTIES OF EQUIPMENT ANCHORAGE

DRAFT Appendix A PROPERTIES OF EQUIPMENT ANCHORAGE INTRODUCTION Based on investigating the potential d;maging effects of high-frequency seismic motions, 1t is concluded that anchorage capacity is the critical structural mode which likely controls the ultimate capacity of nuclear power plant (NPP) components subjected to high frequencies. This is true since displacements at which anchorages fail are relatively small c0111pared to other structural elements. It was conservatively assU1'118d in the high-frequency study that all nonlinear response will be concentrated in the anchorage, and no credit was t,ken for potential nonlinear response in other structural elements. With this ass1111Ption, various anchorage systems were considered. Based on a review of different anchorage types the 1110st critical categories are bolted anchors and welded connections. Equipment with these anchorage systems was investigated further. Electrical cabinets with either bolted or welded anchorage systems were selected as conservative surrogates for all equipment in nuclear power plants. As discussed in Section I, performing the high-frequency study for these systems produced conclusions and recommendations that _are applicable (and conservative) for all equipment. The following sections present the properties that were found for bolted and welded connections, and that were used in the high-frequency study. BOLTED ANCHORAGE Essential electrical and mechanical equipment are co111110nly anchored by using either expansion anchor bolts or cast-in-place anchor bolts. The failure of 1 cast-in-place anchorage is in most cases due to failure of the bolt which represents a ductile failure with force-displac911ent characteristics similar to that of a steel bolt. On the other hand expansion anchor bolts are considered to have brittle failure characteristics since concrete failure ts often their dominant IIOde of failure. Therefore 1n this study the expansion anchor bolts were considered to be the most critical type of bolted anchorage and the emphasis is placed on investigating the force-displacement charactert,tics for these types of fasteners. A-2

DRAfl In recent years there has been a large number of anchor bolt testing programs conducted by manufacturers, users and national lab organizations. After an extensive literature search, a 11st of references pertinent to force-displacement properties of anchor bolts was identified and a review of these documents was performe<I (References l to 8). The review indicates that the ultimate displacement of expansion anchor bolts varies with bolt size, embedment length and concrete strength and is generally larger for shear than tension failure. In addition, it was noted that the expansion anchor bolts have an ultimate displacement capacity generally greater than 0.1 inch for the types of anchor bolt connections encountered in NPP's. Figures A-1 and A-2 show example force-displacement curves for single-direction lo.ading for shell-type anchor bolts for both tension and shear loading. Since the ultimate capacity of the hfgh-freque~cy components is controlled by the anchorage ultimate displacement and the ultimate displacmaent capacity of bolted anchorage is 111t1ch larger than the welded connections (see the following section), ft can be concluded that the bolted anchorage requires a higher ductility scale factor in order to reach failure. The~fore, the welded connections are more critical for investigating the damaging effects of high-frequency seismic motions. WELDED ANCHORAGE The failure of a weld-represents essentially a brittle failure mode, that generally 1s considered to have a ductility scale factor, F,, of 1.0. This is true because small welds used in equipment anchorage have low displacement

• cap,city. For example, a 3/16-inch fillet weld w11l fail in transverse shear when the weld has displaced only about 0.01 inch. For low-frequency motions (i.e., 2 to 5 Hz) once the yield displacement is reached (1.e ** about 0.001 inch) a negligible increase in the earthquake inp~t will cause the weld to exceed the ultimate displacement capacity and fiil. However, as discussed in Section 2, high-frequency components (i.e., greater than 10 hz) can have significant capacity beyond yield. The purpose of this-section is to sUl'lllarize the properties of welded anchorage that were used in the high-frequency analyses.

Fillet welds wen selected since they an the 110st pred011inate type of weld used in the anchorage of electrical cabinets in nuclear powar plants. A literature search was conducted to determine fillet weld properties. Figure A-3 shows scheaatically a fillet weld that is used to describe the pr~perties . . . A-3

12--,---.---,,----,---,--,---,-----,---,----,---.--,---, 10 Z

                      -6 0

(/)

                   ~ +

I-2 0 -+---.---,.----.---.--.-----,----,---,-----.----.-..---i 0.00 0.05 0.10 0. 15 0.20 0.25 0.30 DISPLACEMENT, inches Figure A-1. Example force-displacement curve for 3/4 in. wedge type anchor bolt in tension, f'c=3700 psi (concrete spalling failure mode) (~). 24 20 (/) Q..

               .Y. 16 n::::
               ~ 12 I

(/) B 0 -t---r--ir--~--r-..--,-----,---,----,---.--,---l 0.00 0.10 0.20 0.30 0.40 0.50 0.60 DISPLACEMENT, inches Figure A-2. Example force-displacement curve for 3/4 in. wedge type anchor bolt in shear, f'c*3700 psi (shear failure through threads) (§). A-4

DRAFT

                                         -,,,/J           y r::::~~/                           X Tivoet

• 0.707 t

                                       ~/

Figure A-3. Fillet Weld Schematic following are gene_ral findings:

• Fillet welds loaded in the transverse direction (i.e., x or y-directions in Figure A-3) have relatively high ultimate strength, but low displacement capacity (i.e., c0111pared to fillet welds loaded in the longitudinal z-direction).
• Fillet welds loaded in the longitudinal direction (i.e., z-direction in Figure A-3) have lower ultimate strength, but higher displacement capacity {i.e., compared to fillet welds loaded in the transverse x or y-directions).
• The yield strength, ultimate strength, displacement at yield and displacement at failure all are linearly proportional to the leg size, t. For example, if the ultimate strength of a 3/16-inch weld is 8.3 kips/inch, then the corresponding capacity for a 3/8-inch weld is twice this value or 16.6 kips/inch.
• The initial stiffness of a fillet weld (i.e., at low stress and strain) per unit length is independent of the leg size. For example, the stiffness per inch for a 3/16-fnch weld is the same as a 3/8-fnch weld. However, a weld sized for a given design shear force using a 3/16-inch weld is twice as long compared to using a 3/8-inch weld. Thus, a connection made with a 3/16-fnch weld ts twice as stiff as a connection made with a 3/8-inch weld for the same strength.

The ultimate shear stress of a fillet weld, ~w* is expressed in terms of the ultimate weld tensile stress, ou. The following are typical values of ~w found in the literature: A-5

Loading 01rect1on Lona1tudjn11 Transverse Reference 0.84 Ou 9 0.83 au 10 0.75 Ou 11 The results presented for Reference 11 are based on review and analysis of the data presented fn this reference for weld tests conducted in the longitudinal and transverse directions, separately, and represent mean values. The ratio of the mean values above for the transverse to longitudinal directions is 1.60 for Reference 10 and 1.40 based on an analysis of the data from Reference 11. Another reference gives a ratio of 1.60 (.ll}, Note that in Reference 11 tests were performed for a series of 7 angles from Oto 90 (42 tests total). A regression curve was fit to the data by the authors of Reference 11 that minfm1zed the error over all angles. The following rec011111endatfon 1s given 1n Ref~rence 13 (that 1s based on the tests reported 1n Reference 11) far the ratio of capacity of a fillet weld loaded at an angle 8 to the capacity in the l ong1 tudi na 1 direct 1on, Re,L: Re/L

• 0. 5 si n1*5 8 + 1. 0 (A-1) where 8 is the angle beween the applied load axis and the longitudinal axis of the weld.

For a transverse weld (i.e., 8 equal to 90°), th1s equation gives Rr,L equal to 1.50, which is about 7 percent higher than the value {i.e., Rr,L equal to 1.4) obtained from analyzing the data for the transverse and longitudinal tests separately and computing a ratio as discussed above. Since Eq. A-1 ts based on a regression analysts that considered data at all angles, it 1s not surprising that a ratio based on mean values calculated at 0° and 90°, separately, is slightly different. Another factor which is pertinent to the strength of welds is the difference between n011inal values and actual ultt1111te strengths. In general, strengths from actual test results are about 10 percent higher than nominal values. For example, the 11ini.mum code nominal tensile strength for E60XX electrodes is 60,000 psi with an actual strength closer to 66,000 psi. Reference 14 gives a 118an increase A-6

DRAFT factor of 1.05 with* coefficient of variation of 0.11. From the tests repQrted in Reference 11 the 118an ratio is 1.07. Fabrication tolerances also contribute to the uncertainty in the actual strength. The relationship between force and deformation of fillet welds has been studied and is reported in several publications (ll, ll, ll). These references give equations for different angles of fillet weld relative to the applied load. A recent relationship by Miazga and Kennedy is shown in Figure A-4 (ll). These curves are consisted with the general findings listed above. The curves in tigure A-4 were used as the basis for the weld capacities assumed in the high-frequency study.

             ~

• 1---,...--~-~---,---..----.---...---t a.,, o.,a Q.tD a.ts 0 .,a a.n

                              . ~lljWorillll WIid Illa, 1/1 Figure A-4. Normalized Weld Curves (ill Weld Properties used 1n study The transverse direction ultimate shear stress was selected to be 1.05 ou based on the data in Reference 11. As listed above, this value is at the bottQID of the range of coefficients. SiMilarly, for the longitudinal direction the corresponding value of 0.75 ou was used. No credit ,was taken for the difference between nominal and actual ultimate strengt,h. The curves in Figure A-4 for the transverse and longitudinal directions (i.e., 8 equal to 90° and 0°),

respectively. Relative to capacities based on other sources these values could be

                                       . A-7

20 to 30 percent low. The strength in the transverse direction was checked against the potential for failure in the base metal (i.e., assUTDing A-36 steel). It was concluded that failure would be through the weld. The curves shown in Figure A-4 were used to represent the force-deflection relationships for a fillet weld with E60XX electrodes used in the analyses. Figures A-5 and A-6 show the curves for a 3/16-inch fillet weld for the transverse and longitudinal directions, respectively. Superimposed on each of these curves is a equivalent elastic-perfectly plastic curve that was used in the nonlinear analyses. The location of the yield plateau was selected so that the areas above and below the actual curve are equal. This insured that the energy dissipated in the weld out to failure is the same for both curves. Based on Reference 13 and using the ultimate shear stress relationship given above the initial stiffness values per length of weld are given by the following equations for the transverse and longitudinal directions, respectively: Transverse Pirection Kw

• 124.3 au (A-1)

Lonajtudjnal Djrectjon Kw m 26.l au {A-2) _ for E60XXX electrodes the yield displacements are found to be 0.001 inch and 0.0045 inch for transverse and longitudinal directions, respectively. These values were used in the study. A-8

Data-Based F u ~ a . 4 k

                   ,               ./                       \                    7.5k Force, k
                       ... /~uivalent Elasto-Pfastic Function r
                 /

I/ 7M0~* p7 I o.._....__ 0.001_ _ _ _ _ _ _ _ _ _ _ ....__ _ _ _ _ _ _ _ _ __ j 0.0105 0 0.02 Displacement, in. Figure A-5. Equivalent Elasto-Plastic Function For 3/16-Inch Fillet Weld I-Inch Long Loaded Transversely Using E60XX Electrodes.

                                        .....-....-------~---==:::::;:5.9:::;::::;:::k=:------------.- - - - - - - - ,

sr:Data:--:-_-:-Baaed--:--:F:-un-~--: (,,------:_.,..::-=---------,------~--,..----

                         //~-                                        I                                *- . . . . . . . . ,. .         5.2 k EquivaJent Elasto-Plastlc Function 1./

I ,...__ I .i Force, k f p p I I 0.003 0.064 0 o.a,.

                                                               ~ In.

Figure A-6. Equivalent Elasto-Plastic Function For 3/16-Inch Fillet Weld I-Inch Long Loaded Longitudinally Using E60XX Electrodes. A-9

DRAFT REFERENCES

1. H. Lindquist. Final Report USNRC Anchor Bolt Study Survey and Dynamic Testing. Hanford Engineering Development Laboratory, December 1982.

NUREG/CR-2999.

2. Bechtel Power Corporation. FFTF Report Drilled-In Expansion Bolts Under Static and Alternating Load. January 1975.

3. Wiss, Janney, Elstner Associates, Inc. Draft Of Report Static. Dynamic and Relaxation Testing of Expansion Anchors. January 1981 .

• 4.

5. 6. ITT*Phillips. 3/B" Expansion Anchor Embedded in concrete Masonry Units and Mortars. Westinghouse Hanford Conipany. Test Results on the Dynamic Testing of Expansipn Type concrete Anchors. 1974. ATEC Associates, Inc. Test for Static Tension and Shear strengths of Red Head Carbon Steel Wedge Anchors in c2,s. 3.7. 5.5) KSI Conventional Concrete. Decllll!ber 1985.

7. Drillco Dev,ces, Ltd. A Series of Reports on Prillco Maxibolts.

8. Abbot A. Hank, Inc. Testing Laboratories. KWIK-Bolt Testing Program. Hilti Fastening Systems.

9. J. W. Fisher, et al. "Load and Resistance Factor Desig~ Criteria Connections.* Vol. 104 ST9, Journal of structural Division. ASCE, pp. 1427-1441, September 1978.

10. Static Tensile Strength of Fillet Welded Lap Joints in Steel. International

- 11. Institute of Weld. IIW Document XV-242-68 of International Test Series, 1968. G. s. Miazga and D. J. L. Kennedy. Behavior of Fillet Welds as a Function of the Angle of Loading. Structural Engineering Report 159, University of Alberta, Canada, March 1986.

12. Kulak and Timler. "Tests on Eccentrically Loaded Fillet Welds.* D~part111nt of Civil Engineering, University of Alberta, Edmonton, December 1984.

13. D. F. Lesik and D. J. L. Kennedy. Ultimate Strength of Eccentrically Loaded Fillet Welde<l Connections. Structural Engineering Report 159, University of Alberta, Canada, Hay 1988.

14. B. Ellingwood, et al. "Development of a Probability Based Load Criterion for American National Standard ASS." NBS Special Publication 577, National Bureau of Standards, June 1980. *

15. Butler, Pal and Kulak. Eccentrically Loaded Weld Connections,* Journal of the Structural Division, ASCE, ~Vol. 98, No. STS, Hay 1972, pp. 989-1005.

DRAFT Appendix B DEVELOPMENT OF SIMPLIFIED DYNAMIC PROCEDURES (2-DOF}

DRAFT Appendix B DEVELOPMENT OF SIMPLIFIED DYNAMIC PROCEDURES (2-0CF) INTRODUCTION

• This appendix gives the general derivations for the simplified dyn1111ic procedures based on the two-degree-of-freedom (2-00F) sliding and rocking 110dels. The simplified procedures were calibrated with the results from nonlinear time history analyses as discussed in Section 4. In Section 3 of this report the equations for the single degree-of-freedom sliding and rocking 110dels are derived.

The following two sections give the derivations for the pertinent equations using the equivalent pseudo linear-elastic approach for the two procedures. The equations are st.111111arized for both sliding and rockin~ models, and an example calculation is given for each case. The two sections are each self contained so the reader can study the derivation for the rocking procedure without first reading the derivation for sliding procedure, if desired. SLIDING MODEL Figure 8-1 shows the proto~ype and 110del for component sliding. In Figure 8-la a squat component with a low aspect ratio is shown. This model can translate horizontally without rocking. It 1s resisted by the welds at the base and the friction force between the base and the supporting surface. Figure 8-lb idealizes the component and defines the pseudo linear-elastic model properties used in the formulation that is presented below. The model consists of two-degrees-of-freedom since a part of the mass is located at the equipraent base. This model captures both the first dyn1111ic mode and rigid body response of the component. Figure B-2 separates the total model response at ult1~ate capacity into equivalent linear dynamic and rigid-body 110des. Note the second mode frequency is sufficiently high that the 110dal response follows the ground motion directly (1.e., rigid-body response). In figure B-2 the total horizontal force 1t the bas-e F,c (that is composed of both the yield force in the weld and the force due to friction) is equal to the dynamic 1110de force, Fs:o, "plus* the rigid-body mode force, F811

• Note that the procedure for combining modes is discussed below.

8-2

DRAfl l,l: T a c l l ~ li'f:=-- Kw:Wlldlll!I-.

                                                                       -=
                                                                       +:

Jt:

                                                                           =--

f'llldtlon . . . . . Bue

                                                                           ~

d Alnt Dlnl!li'IGRllllloat FIDd-8IN ~ II : Aalllnllal1 Clue to cn.t:y L ~ (Sqtiat) b. llllntModtl Figure B-1. Model for Component Sliding Other variables in Figure B-2, that are not defined in Figure B-1, are as follows: W: Total weight of the model (1.e., Mg} Displacement of top mass at ultimate response Displacement of bottom mass at ultimate response (i.e., the displacement limit of the weld} Secant circular frequency corresponding to ultimate response, which is equal to 2sf,, where f, is the corresponding secant cyclic frequency Z: Support motion zero period peak acceleration Fraction of total mass missing from dynamic mode (also fraction of total mass in rigid-body 110de) r: Dynamic mode participation factor Spectral acceleration capacity at secant frequency, f,, and secant damping, ~. Figure B-3 shows the force deformation diagram for the weld spring. Properties for the weld spring are discussed 1n Appendix A. Note that the component spring stiffness,~. is assumed to remain elastic. The following are definitions of tenns used in the formulation of the equations for determining the sliding 110del response. B-3

11111 W

                                                                 .._2 (1-*>g &tu
                    **>W D
                                                              *rSae (1- er)~ ~tu m                                                                           D I
                                                                                 +

IatallladlfBNPQOM BlgkHkldJ Madi Figure B-2. Sliding Hodel Modes at Ultimate Response

Fy : Yield Force 6y: Yleld Diaplacement

                                              &u : Uhlmate Displacement 6y                             ou Figure B-3. Force-Displacement Diagram for Weld Fixed-base frequency, f,:

(B-1) Normalized secant frequency squared, X: Nonlinear un,tless factor, A: (B-2) (B-3) Weld yield capacity coefficient, Cu: Cu* F/W (B-4) Normalized total base force, F1JW: F _!:

• t + Cu (B-5) w B-5

DRAFT Nonnalized dynamic mode base force, FaofW: Fao Fac/W w -I\.- (B-6) where~ is the ratio of the total base force to the dynamic mode base force This term is discusseu below. Relative spring displacement, 6r (B-7) From Figure 8-2 equilibrium between the inertial fo'rce at the top mass and the spring force gives the following equation: (B-8) Combining Eq. 8-8 with Eqs. 8-1 and 8-2 leads to: (B-9) Using Eq. 8-7 it can be shown that: 6 - r _x_ 1 - X 6 u (B-10) and H., - - 6 1 (B-11) u 1 - X u Normalized secant freauencv sauared, x The following steps lead to the equations for the normalized secant frequency squared, X. Using the dynamic 1D0de in Fig. 8-2 and equating the forces at the base the following equation is obtained: (B-12) From Eqs. 8-3 and 8-11, Eq. 8-12 can be expressed as follows:

    << A X2   - (A+ 1) X + l

• 0 (B-13) 8-6

Solving for X in terms of A and* the following expression for Xis obtained: X ., 2 (B-14) (1 +cos8}{1 +A) where: 2 a. A B= (1 + A) 2 The secant frequency, f 1 , is just: (B-15) Fraction of Mass Missing from Dynamic Mode The fraction of mass missing from the dynamic mode, F", is equal to the mass in the rigid body mode divided by the total mass. Following the procedures for linear elastic dynamics F" is given by the following expression: w w (1-a.)- (l -r6~) +a- (l -r6u) g 9 (B-16) F" * - - - - - - - - - - - - - W/g or simply: (B-17) The relationship for the dynamic mode participation factor, r is as follows: (1 - a) 6~ + a&u r- - - - - - - - (B-18} (I - a) &t,} + at,} Using Eqs. B-18 and B-11 and rearranging the terms: 1 + (-a-)(1 1 - a

                                  - X) r 6tu   - --------                                                       (B-19) 1 +  (-a:-)

1 - ex (1 - X) 2 and B-7

DRAFT (8-20) Substituting Eqs. B-19 and B-20 into Eq. B-17 and simplifying, FM becomes: ciX2 (B-21) Spectral Acceleration capacity

• The equivalent linear elastic spectral accelerati~n capacity is expressed in terms of an effective dynamic mode frequency, we, that lies inbetween the fixed-base frequency and the secant frequency. As the weld becomes nonlinear the frequency decreases and the model responds to spectral input at a lower frequency. Fig. B-4 shows the forces for the dynaaic mode that have been adjusted to convert from the secant frequency, ~s to a higher effective frequency, w** In order to maintain equilibril1111 the dynamic force at the base must be increased_by the ratio ~!f~! as shown. The need for an increase in the base force is discussed in Section 3 (e.g., see Figure 3-1). Fr011 Fig. B-4 equilibrium of the base and inertial forces gives the following equation:

rSae w (1

• a) -g &lu f-----1 rs*** wi au Figure B-4. Sliding Hodel Dynuiic Mode Forces Adjusted to. Effective Frequency (B-22)

B-8

DRAFT where: sa.

• spectral acceleration capacity at the effective frequency, f., and effective damping, ~ *.

Defining a normalized effective frequency squared, x.:

                "\2      f2 X * --- or      _e_                                                 (8-23)
          *   '    2        2
                ~        ff and using the definitions for r &tu and r gu fr011 Eqs. 8-19 and Eq. 8-20 and F" from Eq. 8-21, Fm/W (Eq. 8-22) is simplified to the following equation:

Feo

      -w    -                                                                (8-24)

When the weld properties and coeffi*cient of friction are known Eq. 8-24 can be rearranged to obtain the effective frequency spectral acceleration capacity, Sa.fg: (8-25) Ratio of pynamic Mode Base Force to Total Base Force The following steps lead ~o the expression required to evaluate the term flu, that was introduced in Eq. 8-6. From Fig. 8-2 and equilibrium for the rigid body mode the equation for-FIIR/W follows: (8-26), where Z

• support motion zero period acceleration (consistent with sa.)

If the ratio S 1s defined to be: X Sa_ s--- - (B-27) X. z Then F,.;w can be express~d in teras of the following equation: B-9

DRAFT For the case where the effective frequency. f 1 is less than the frequency corresponding to the peak of the response spectrum, the *best estimate* coatlination of the dynamic and rtgtd-body IIIOdes ts obta1ned by th~ square root of the sum of the square root (SRSS) procedure. C0111b1n1ng the dynamic and rigid body base forces by an SRSS combination, the following equation ts obtained: (B-29) Substituting Eqs. 8-24 and 8-28 into Eq. 8-29, FmfW becomes: (B-30) Comparing Eq. B-6 with B-30 it ts seen that Ru ts just equal to the denominator on the right side of Eq. 8-30. For cases where the effective frequency is above the frequency at which the spectral acceleration is well below its peak, the dynamic "best-estimate* combined response is obtained when the rigid body 110des are combined algebraically. Inbetween these two frequencies a linear interpolation can 1,e, used. This leads to the following definition of Ru: (B-31) where: 8-10

DRAFT (SRSS cOlllbination)

                   ~   -                               (Algebra*1c cDllbination) f P

• Frequency corresponding .to the peak of the response spectrum fh
• Frequency higher than f 11 at which the spectral, acceleration is well be1ow the peak of ttte response spectrum Table B-1 gives estimated fP and fh values for the earthquakes used in the study.

As explained below, Ru in general is not known a priori since Sas given by Eq. B-27 is not known. This requires an iterative procedure to find the ductility scale factor. Effective Damping Ratio The damping ratio at the maximum displacement is determined first and 1s converted to the damping ratio at the effective frequency, f 8

• Damping Ratio at the Maximum Displacement. The damping ratio at the maximum displacement, P., is given by the following equation (1):

(B-32) where AE 1

• Elastic energy dissipated in one full cycle of pseudo linear-elastic dynaaic mode AEz *.Hysteretic energy dissipated in one full cycle of pseudo linear-elastic dyn1111ic IIOde E
• Total energy in pseudo linear-elastic dynamic 1110de at the secant frequency The elastic damping, P1 , is defined in terms of the fixed-base frequency, f 1* The elastic energy in one cycle is just:

21' (B-33) B-11

DRAFT Table B~l CUTOFF FREQUENCIES USED IN STUDY fp f ' H2.... Eutbgy~~!! .wh il!tl I R.G. 1.60 (Artificial) 2.6 15 2 Olympia, WA 4.8 7 3 Parkfield, CA 1.6 4 4 Tabas, Iran 5.3 9 5 IIIIJ)erial Valley, CA 2.8 12 6 Nahanni, Canada 5.6 20 7 Saguenay, Canada (Site 20) 7.1 12 - 8 9 10 11 Gazli, USSR (East) Bear Valley, CA Gazl1, USSR (North) Saguenay, Canada (Site 16) 7.5 5.6 13.3 13.4 23 20 36 7 12 Leroy Modified 17.0 19 13 Leroy, Ohio 21.6 24 14 New Brunswick, Canada 20.6 55 15 Artificial 21.6 55 8-12

where: (component spring) (weld spring) (component spring} (weld spring)

                ~    =  tangent stiffness of anchorage It is assumed that the elastic damping in the weld spring is effective up to the yield displacement after which it drops to zero (i.e. tangent stiffness damping).

This same assumption was_made in the nonlinear time history analyses perfonned using DRAIN-20. Thus tE 1 , can be written: {B-34) where: (see Figure 8-5)

                 &Y

• weld displacement at yield Noting that~ &Y is equal to Cu Wfort less than ty, Eq. B-34 is evaluated and becomes:

{B-35) B-13

DRAFT dlsplacement, X 1 -1 1 t --sin - Y s 11w time, t-ty 'It 2s Figure B-5. One-Quarter Cycle of Displacement in Weld at Ultimate Response where: C Z -1[ µ,.

                      * -;2 µ,.sin     i]    +   rz-:

yµ,. - 1 The hysteretic damping is due to friction between the base and support, and yielding in the weld. Figure B-6 shows the 11aximum hysteretic loop in the weld in terms of the 12W force at the base. A fraction of the base capacity is associated with the rigid-body mode; thus, only a portion of the hysteretic energy is assigned to the dynamic mode. This is done by assuming that the force F10 that is equal to (t+(:u)W/ffu is effective for the portion of energy assigned to the dynamic mode. With this IIIOdification AE2 1s obtained fro111 Figure B-6: (B-36) The total energy in the pseudo linear elastic dynamic IIIOde at the s,cant frequency is just: E

• 2 -W2[

                 <il (1 - a) ~t.;2 + ati;;2]                                     (B-37) g B-14

F m I U'I Figure B-6. Hysteret1c Loop in Sliding Model Weld

DRAFT Substituting Eqs. B-35 through B-37 into B-32, the expression for P. becoaes: (B-38) Making substitutions from previous relationships that relate the frequencies and displacements to normalized terms, P. can be put into the following form: (B-39) where: D

• 1 + _<<_ (1 1 - Cl Cu_

G "' Equation B-39 can be expressed as the sum of two terms: (B-40) where: (B-41) (B-42) (B-43) For more realistic partial cycles the hysteretic damping tenn PH can be estimated from b~h* where bis an empirically-obtained coefficient between O and 1.0. the term: 1 - 6 C) X112 P., ( 1 - G) is generally relatively small and can be neglected. In addition, G can be approximated by 1/µw without significant loss of accuracy. This leads to the B-16

DRAFT following expression for ~h:

        ~h

• 3_

1t (1 - _!) D (1 - G} (B-44) Eq. B-43 was used in the calibration analyses discussed 1n Section 4. However, for practical purposes Eq. 8-44 can be used when applying the silllJ)lified methodology to reduce evaluation response spectra. Damping Ratio at Effectjye Frequency. To convert P. to P. the dissipated energy in a full cycle must be kept constant. Thus following the form of'Eq. 8-38, the expression for P. and P. are as follows: M, + ~ (B-45) 2 ~ ci [(1 - <<) ~~ + <<~] g M, + M2 (B-46) Therefore the relationship between P, and p1 is simply: (B-47) or (B-48) where bis a calibration term as discussed below (i.e., see section "Determination of Effective Frequency and Damping"). Puct1Jttv scale Factor The ductility scale factor, FP, is the ratio between the earthquake level that causes failure and the earthquake level that causes yield. The ratio can be defined between peak ground acceleration values or between corresponding elastic response spectral ordinates (i.e., at the same frequency and daaping ratio). The reference yield level is defined in this report as the earthquake level that causes the fixed-base model (i.e., no base-displacement} to reach the yield .f!2.o:i in the weld spring. At the yield level the sum of the inertial forces in the two masses exactly equals the sum of the weld spring force plus the maxilllUII friction force at the base of the IIIOdel. B-17

This definition is contrasted to a 110re c011110n definition that is soaetimes used where the yield level is defined as the earthquake level that causes the weld spring to reach the yield displacement. The definition used here is preferred for the following two reasons:

• Structural analysts usually do not include the flexibility of welds in calculating a component frequency. Thus, defining the yield capacity at the fixed-base frequency is consistent with design practice.
• The yield displacement is negligibly small (e.g., 0.001 inch for a 3/16-inch weld size loaded in transverse shear);

therefore, the difference between yield capacities based on the two definitions is small. The ductility scale factor can be expressed in terms of a reference earthquake level characterized by a family of elastic response spectra at different damping values. {B-49) where: Factor to scale the reference spectra to the level corresponding to failure (1.e., displacement in weld equal to the ulti1111.te limit) Factor to scale the reference spectra to the level corresponding to yield (i.e., force in weld of fixed-base model is equal to the yield force plus the IIIX1mwn friction force) The FSu factor is just equal to the ultimate spectral acceleration capacity sa., divided by the reference spectral acceleration at the effective frequency and effective damping, Sa(f., Pe>* From Eq. B-25 the spectral acceleration capacity can be expressed 1n terms of the aodel properties. Noting that Fia/W 1s equal to ( + Cu)/1\i from Eqs. B-5 and B-6, FSu bec011es: FS * + Cu X. g (B-50) u Sa{fe, 13,,) X Ru(l - FM) B-18

The expression for FSY can be derived in I similar inner as FSu starting with a free body diagram. An alternate way to obtain FSY is to note for the fixed-base model with a frequency ff and dlllll)1ng Pf t~at there is no frequency shift. Hence, X ind x. both equal to 1, and FM is just equal to* froa Eq. 8-21. Hence using Eq. B-50:

                  +Cu          g FSy *                                                                     (8-51)

Sa(f,, ~) R,(l - Cl) where Ry is obtained from Eq. B-31 with FM equal to 1, and S is equal to Sa(ft, Pf) A divided by the corresponding peak ground acceleration Z. Finally, combining Eqs. ~ B-49, B-50 and B-51: (B-52) Determination of Effect1ye frequency and Damping In developing the pseudo linear-elastic model as presented above the concept of an effective frequency, f., and effective damping, P., are introduced. Since the. model frequency and damping change as the weld becoaes nonlinear, it is necessary to use effective or averagen values which are in between the fixed-base and 0 secant model properties. It 1s through these two variables where the model is calibrated to fit*the results of the inelastic DRAIN-20 analyses. To this end the following empirical re)ationships are defined: 1 - x. * (I - X)

                       *                                                          (B-53)

(B-54) where Ph is the aaxilllUIII possible hysteretic damping at the secant frequency (i.e., maximum displacement) giv~n by either Eq. B-43 or Eq. B-44. As discussed in Section 4 of the report, best fit values for I and b were found to be 1.6 and 0.3, respectively. Procedure for Calculating Duct111tv Scale Factor The ductility scale factor, F,, is calculated following the equations given above. For the case where the weld yield force is known, iteration 1s required due to the tenn flu (see Eq. B-6), *which is the ratio of the total base force, Fae, to the B-19

dynamic mode base force, F., (see Figure 8-2). This is because Eq. 8-31 for Ru depends on the ratio S (see Eq. B-27) that in turn depends on the spectral accele~ation capacity at the effective frequency and damping, Sa(f., ~.). The latter teni is unknown a priori. When the weld yield force is known, Ru converges very quickly. The more general problem is the determination of a reduced response spectrum Sar as discussed in Section 2 of this report. For this case both Ru and Sar are unknown and two levels of iteration are required. However, in practical problems the two iterations are combined into a single process where both parameters are corrected simultaneously in each iteration. The steps for calculating F, using the simplified 11ethodology for the sliding model are given. in Tables 8-2 and 8-3. Here the equation for Cu, the weld yield capacity coefficient, ts written in terms of the reduced spectral acceleration, Sar. Using Eq. 8-51 where FSY is equal to F111 {i.e., required minimum safety factor) and Sa(f1 , ~f} ts equal to Sar* (t + Cu} is found directly to be: ( + Cu) = F'" sa.. Ry (1 - u) (B-55) g For cases where the yield capacity, Fy, for the weld ts known, then Cu is just: cu - F.;w (B-56) F_igure 8-7 shows a flow chart for calculating the response spectrum reduction factor, which is just equal to 1/F, (see Section 2). This figure is applicable for both the sliding and rocking lllodels (see next section). Example ca1cuJat1on figure B-Sa shows an example 5-percent damped unifor11 hazard spectrum for a EUS site *. Figure B-8b shows the corresponding reference spectra for different damping values. For this exaaple the reduced *s-percent d1111ped spectral acceleration ordinate, Sar, at the fixed-base frequency of 15 Hz 1s calculated for three trials is tabulated in Table B-4. Note that the aodel parameters assumed in the analysis are provided at the bottom of the table. Initially 1n trial 1, values of 0.507g and 1.0 were assumed for Sar and Ru, respectively. The calculation steps 1n the flow chart shown in Figure 8*7 and the equations given in Tables B-2 and B-3 were followed. At the end of each trial new values of Sar and Ry were calculated that were used in the subsequent trial. By the third trial convergence was obtained. 8-ZO

UAAfi Start with original (ln'8dualld) Sa1 value [I.e., Sa(l f. ~, )J Aawne Reduced Fleaponu 8')eoln.m Value, S.r Aalume Dynamlo Fon>> Ratio, Ru Calcutme Oudfflty Scale Faca, F"' Siding Modal: Tlblt 8-2 Roddng Modet Table 8-5 Calculate New Sar (l,e., Sar) aa;

• Sat JF11 C8lculate New Ru (Le., R~)

SeeTe!M B-3 No Figure B-7. Flow chart for calculating response spectnnn reduction factor, SaJSa1 , using s1mpl1f1ed aethodology for sliding and rocking aooel s. 8-21

DRAn 1 ~ ' si

 ~      l I                                                                        "-----0.11 i
        +1-~--.--~-*-0,--.0IS--,.,--.-,--.-,~,~,~,~1'0_ _ _20____--,~.....--..-,...~
        ~

0.1

                                                                     .,_,             I   *100 Frequency' f, (Hz)

a. Evaluation Response SpectrUIII, Sa, 1.00 .-----,...,..-rril'TTT--,--,--,-,-"'!"T"I~-..--.,....,...,..,...,.........

0) 0.90 p z Q0.80 I- _JJ_

        ~ 0.70                         0.0-4 w                               0.05 0.06
       ...J 0.60 w                               0.08

o. 10 u o.50 0.12 0 0.15

        <(                             0.18 0.40                    0.21

_J 0.25

        <(

O::'.'. 0.30 1-u wo.20 Q_ CJ) 0.10 1 10 100 FREQUENCY, hz

b. Evaluation Response Spectra at Different Damping Values Figure B-8. Response Spectra Used the Example Calculations B-22

Table B-2 SIMPLIFIED METHODOLOGY STEPS FOR SLIDIN6 HOOEL Step Equation Constant NorH]ized Effgctj~~ ~ ° COIIJ)Onent fixed-base Freouency stiffness (2

• ff)Z&ufg

1. Basic Hodel Paraaeter, A A "' Mm Component mass

( + Cu)/R,i

• s Fraction mass at base 6u
• Weld ultinate where: ff - 1~

2* (1 - <<}H g 0 displacement capacity Acceleration due to a, I Sa,. gravity N w ( +Cu)

                                               -       FSM-
                                                             ,9 Ry (I - 11) s~     Reduced spectral acceleration

2. Nonnaltzed Secant Frequency Squared, X X - 2 (1 + c~8)(1 + A)

F9'1

• Required min1111t111 safety factor.

where: sin-I .fie ~

• Obtained fr011 previous iteration (see Note 1) 8 C Ry
• See note 2 B . ,2 << A (1 + A) 2 a Coefficient of sliding I

friction a a EIIJ)irical frequency

3. Normalized Effective I - X8 * (I - X)a constant (dfaensionles*sJ. A Frequency Squared, 1e value of 1.6 was found to be aooropriate (see Section 4)

Table B-2 (Continued) SIMPLIFIED HETHOOOLOGY STEPS FOR SLIDING MODEL Step EQuation Constant Eff~t1~~ 0~1!!1!11!!1 B~tjg

4. Effective Elastic Daaping Ratio at Ultiiute Drift,

                                  ~fp  m
                                              ~                                      pf - Elastic damping ratio at fixed-base frequency ca Ptp b  - constant Empirical  daaping (di11ensionless).

N I .... 5. Hysteretic Damping Ratio at Ultimate Drift, Ptt ~ "' b  ! (1 -~) [1 - ~] A value of 0*.3 was found to be appropriate (see Section 4) D )(1 -+ (- 11 1 - Cl X) 2

                                                                                             &u
                                                                                     ~  "'
                                                                                             &y 6y

• Weld yield displace11ent

6. Effective Du,ping Ratio at.Effective Frequency, ~e Pe - -

X Xe (Pfp + I\J)

Table B-2 (Continued) SIMPLIFIED METHODOLOGY STEPS FOR SLIDING MOOEL Step Eauation Constant Du~t1ljtr s,11~ E1,tgr

7. Fraction of Total Mass Missing From Dynu1c FM - I [(r.J1 -x)2j

                                                           +
                                                                .x2 Model, FM tD N'

U'I

8. Ductility Scale Factor, FIi FP * [ x.X]['** ~, ][~

Sa(fe, Pe HRy l 1 - fft Ru Sa(f, P) = spectral 1cceleration at frequency, f, and da111ping, , where: fe = ff {x; fll!tt: I. I\, detemined from Table 8-3 using f .. f8 and P * ~e (Note that R,, is iSSlllll!d tn1t1ally for the first trial and revised values are obt1ined fr011 Table 8-3 for subsequent iterations.)

2. '\, determined fr011 Table 8-3 using f ., ff, ~ * ~f' _!_ = 1 , and FH = 1 Xe

Table 8-3 DETERMINATION OF RATIO OF TOTAL BASE FORCE TO DYNAMIC MODE BASE FORCE, R, AT FREQUENCY, f, AND DAMPING RATIO, p (USED TO OBTAIN '\J AND R_y) Frequency Range1 r?- Equation3 f s fp R * ') 'P C ' I+[ (I fa

                                                                                                              - Fff)S r

f ~ fh R* f\i f\i

• 1 + FM (1 - f°H)S tl:I

                                                                                                                    -X Sa(f, ~)

N I R .. ( f - fpJ {Rti - Rp) where: s Xe fp .S. f .S. fh C °' Rp + (fh - fp) i z.. Peak ground acceleration consistent with Sa(f, P) FM, X, Xe, Sa(,) = see Table B-2 (Sliding) see Table 8-5 (Rocking) Frequency below which dynamic and rigid body responses are combined by SRSS (i.e., peak of response spectrwn) Frequency above which dyna~ic and rigid body responses are comined algebraically For rocking 110del FM fs equal to -Ff,E

Table 8-4 EXAHPLE SLIDING HODEL ANALYSIS A B 6 D I 0.507 1.0 0.615 0.374 0.079 0.4m 0.7!B 0.898 14.212 1.014 0.044 0.043 0.073 0.114 0.444 1.231 0.412 1.005 2 0.412 1.005 0.500 0.462 0.<Bi 0.429 0.716 0.867 13.965 1.020 0.042 0.051 . 0.077 0.101 0.434 1.263 0.401 1.(XM 3 0.401 1.004 0.487 0.474 0.lll7 0.431 0.711 0.ffi3 13.933 1.021 0.042 0.052 0.078 0.099 0.433 1.269 0.400 1.004 tD I N N:xfe1 ParaEters (Permissible 1rtelastic anchor d1splaam!lt

• 0.01 inch)

F,i

• 1.5 Sai
• Sa(ff> ~f)
• 0.507 g fp -= 25 Hz 3u .. 0.01 inch Z* 0.46 g fh _"! 33 Hz II
• 0.2 I\,
• 10 R_r
• I.OH ff
• 15 Hl Pr
• o.05
• 0.0

Thus, the evaluation spectrum at 15 Hz is reduced from S1 1 equal to 0.507g to a value of Sar equal to 0.400g. ROCKING MODEL Figure B-9 shows the prototype and 110del for ctmponent rocking. In Figure B-9* a tall component with a high aspect ratio is sh0wn. This IIOdel can rock without translating. Rocking is resisted by welds on each side at the base and the vertical gravity force. Figure B-9b idealizes the component and defines the pseudo linear-elastic model properties used in the fonnulation which is presented below. The model consists of two-degrees-of-freedom because of the rocking and translation modes of response. This model captures both the first dynamic mode and rigid-body response of the component. Figure B-10 separates the total ~odal t II: Tlllill~.._ Kr:~....,._..,_ b

              .,_,.                                        Kw: WIiii~
                                                            .:    ~     ......_

t : . . . ~ HliQl'II

                                                                 .,__.._,.._OI _

th ~

                                                            -,,c.-w--~1'111111
                                                                            ~
                                                            ' : ~ 1M. Gil'M4ly y

I I 1'1111 llabl:8-~ ...

                                                                   ,.._...IClr_,.,llel;IIIII w

tc Ill L~(fllll) Figure 0-9. Hodel for C011p<Jnent Rocking response a~ ultimate capacity into equivalent linear dynamic and rigid-body aodes. Note the second mode frequency 1s sufficiently high that the modal response follows the ground motion directly {1.e., rigid-body response). In Figure B-10 the total vertical force at the base Fae (that is coaposed of both the yield force in the weld and the force due to gravity) 1s equal to the dynamic mode*force. F111 , "plus" the rigid-body 110de force, F111

• Note that the procedure for combining modes is discussed below. Other variables in Figure B-10, that are not B-28

(1 * *) M

                                                                 .,._,. t1 **>;     ltu m                                                             ~ r Saa    (1 * *> !II *tu N

u,

                                                                                        +

Tlllllllllllla... lllillll:IAdi, mall Figure 8-10. Rocking Modes it Ultimate Response

                                                                    -'riAFl defined in Figure 8-9, are as follows:

W: Total weight of the model (i.e., Mg}

    &tu: Horizontal displ1ce11ent of top mass at ultimate response 6u: Vertical displacement of weld at ultimate response 6r: Relative spring displacement e: yh/w
     ~.=   Secant circular frequency corresponding to ultimate response, which is equal to Zif,, where f, 1s the corresponding cyclic secant frequency Z: Support motion zero period peak horizontal acceleration F11 : fraction of total mass missing from dynamic mode (al so fraction of total ma~s in rigid-body mode) r: Dynamic mode part1c1pat1on factor Mass moment of inertia about the corner less the Mass moment of inertia of the.)op and bottom mo~el masses about the corner (i.e., less M[ /4 + (1 - a) {yh) ]). For uniformly distributed mass [i.e., y

• 0.5/(1 - a)], AM 0 is given by the following expression:

Sa,: Spectral acceleration capacity at secant frequency, f,, and secant damping~. Figure B-3 shows the force deformation diagram for the weld spring. Properties for the weld spring are discussed in Appendix A. Note that the component spring stiffness, Kf, is assumed to remain elastic. The following are definitions of terms used in the formulation of the equations for detennining the rocking model response. Fixed-base frequency, ft: (B-57) Normalized secant frequency squared, X: B-30

(B-58) Nonlinear unitless factor, A: (2 1t ff) 2 au (C. + f 1 )/g A * (B-59) FaofW where:

c.

• e 2 ( I - a) fl = (0.25 + Al\)

w2H

                        .!.  [1  + e 2 (1 - u) (1 - 4u)] (for uniformly distributed mass) 3 Weld yield capacity co~fficient, Cu:

Cu

• F/W (B-60)

Nomalized total base force, F9JW: Fae

         -

• 0.5 + Cu (B-61) w Normalized dynamic mode base force, FaofW:

(B-62) where Ru is the ratio of the total base force to the dynamic mode base force. This term is discussed below. Relative spring displacement, 6r (B-63) From Figure B-10 horizontal direction equilibrium between the inertial force at the top mass and the spring force gives the following equation: (B-64) Combining Eq. B-64 with Eqs. B-57 and B-58 leads to: B-31

(B-65) Using Eq. B-63 it can be shown that:

           .      X 6 * --liue                                                         (B-66) r  1 - X and (B-67)

Normalized Secant Freauency sauared, X The following steps lead to the equation for the normalized secant frequency squared, X. Using the dynamic mode in Figure B-10 and taking moments about point a at the base of the model the following equation is obtained: (B-68) where: (0 is the angular acceleration) Note that the effect of gravity is included in the term F80 (see Eqs. B-60, B-61 and B-62). Substituting in the expressions for 6 and f 1 , Eq. B-68 becomes: (B-69) Then substituting in the expression for 6tu and Ce, Eq. B-69 becomes: (B-70) Next, the expression for A (1.e., Eq. B-59) is introduced and Eq. B-70 is rearranged to: B-32

r: ~ t.. ... *

                                                                            \\I J-        ::

[_f_, -] A X 2

                        -  (A + 1) X + 1

• 0 (B-71)

c. + f, from which X can be solved in terms of A and f 1/(C. + f 1):

2 X* (B-72) (1 + cos8) (1 + A) where 8 * (1 + A) 2 the secant frequency, f., is just: (B-73) Fraction of Mass Missing From Dynamic Mode The fraction of mass missing from the dynamic mode, F", can be derived with assistance from Figure B-11. In Figure B-lla the forces for the dynamic mode are - shown in terms of F". In Figure 8-llb the force and moment are shown in terms of the dynamic mode properties. Solving for the force, F, in the weld by taking m0111ents about point a for both free body diagrams and equating the results leads to the following expression: (8-74) Using the relationships for f 1 and e and solving for F" Eq. 8-74 simplifies to: r& f (1 - r &1u) - u I (B-75)

                                          - ( 1 - a:) e B-33

                    ----J Sa 8 ~ (1 - a) (1 - Fy)                                   ~         r atu aa_ w i (1 - *>

!XI I .,..w F F

b. Forcn and Moment In Terms of Modal Propertln

                                     .,1   .

Figure 8-11. Dynamic Mode Free Body Diagrams For Determining Fraction of Hissing Mass, FH

DRAFT The relationship for the dynamic mode participation factor, r, is as follows: (1 - 11) 6\i + 0 r = (B-76) o - u) 6~z + &\ fl [ M + w2] 4 where: e ., &u w Introducing the expressions for f 1 , e and Ce and rearranging the terms in Eq. B-76: (B-77) and (B-78) where: Substituting Eqs. B-77 and B-78 into Eq. B-75 and simplifying, FM becomes: (B-79) Note that FM is always a negative value which indicates that modal mass in the dynamic mode always exceeds the total mass. F111 1s defined as the fraction of excess mass, i.e., (B-80) B-35

Spectral Acceleration Capacity The equivalent pseudo linear-elastic spectral acceleration capacity is expressed in tenns of an effective dynaaic IIIOde frequency, a8 , that lies inbetween the fixed-base frequency and the secant frequency. As the welds become nonlinear the frequency decreases and the model responds to sper.tral input at a lower frequency. Figure 8-12 shows the forces for the dynamic mode that have been adjusted to convert fr011 the secant frequency, w., to a higher effective frequency, w** In order to maintain equilibrium the dynamic vertical force in the weld at the base must be increased by the ratio ee 21~.2 as shown. The need for an increase in the W 6u A. rSa 8 - - II g 2 6tu

                        ~r yh

-

• w ll>e2 Fao

                        ~

Figure 8-12. Rocking Hodel Dynamic Mode Forces Adjusted to Effective Frequency base force is discussed in Section 3 (e.g., see Figure 3-1). From Figure 8-12 equilibrium of 110111ents about point a gives the following equation: 8-36

I., '

                                                                                                          ~

_,,; /I -

                                                                                      .     *4-t:., :
                                                                                                   ~
                                                                                                     ~(

where: sa. - spectral acceleration capacity at the effective frequency, f. and effective damping, Pe** Defining a normalized effective frequency squared, x.: f2 or -*- {B-82) f' f and using the definitions for Ntu and r,\ from Eqs. 8-77 and Eq. B-78, a1ong with

• the expressions for' f 1 , c., and D the following expression for f.lW is obtained from Eq. 8-81:

F10 X Sa_

          -    .. -   -     (I + FIi!) (1 - u) e                                    (B-83}

w ~ g When the weld properties are known, Eq. B-83 can be rearranged to obtain the effective frequency acceleration capacity, Sa/g: Sa_ ~ Fer/W l

          - g X ( 1 + FME) ( 1 - u) e (B-84)

Rat10 of Dynamic Mode Base fac'tor to Total Base force The following steps lead to the expressions required to evaluate the term I\,, that was introduced in Eq. B-62. From figure B-10 110111ent equilibrium for the rigid body mode leads to the equation for FufW as follows: Fu Z

               * - F11 e ( 1 - u)                                                  (B-85) w     g where:

i

• support motion zero period acceleration (consistent with sa.)

If the ratio Sis defined to be: XS~ (B-86) S *- - X. i B-37

DRAFT th!n F1JW can be expressed tn terms of the following equation: F11 . x Sa. F"

          -     =-    -     - e    (1 - 11)                                (B-87) w      X. g s For the case where the effecttve frequency, f. is less than the frequency corresponding to the peak of the response spectrWII, the *best estimate*

combination of the dynamic and rtgtd-body modes ts obtained by the square root of the sum of the square root (SRSS) procedure. Combining the dynamic and rigid body base forces by an SRSS combination, the following equation is obtained: (B-88) Substit~ttng Eqs. B-83 and B-87 into Eq. B-88, Fai/W becomes:

            =  -;:::::=====                                                (B-89) 1+        FMl

[ (1 +F11;)S r Comparing Eq. B-62 with B-89 it is seen that Ry ts just equal to the denominator on the right side of Eq. 8-89. -- For cases where the effective frequency is above the frequency at which the spectral acceleration ts well below its peak, the *best estimate" combined response ts obtained when the dynamic and rigid body modes are combined algebraically. Inbetween these two frequencies a linear interpolation can be used. This leads to the following definition of Ry: I\, f. i fp Ry - ~ f* ~ fh (B-90) I f _fl I\,+ [ _*_P fh - fp (I\, - I\,) fpif* .s_fh where: B-38

DRAF* 1 + [ (1 + FME FIU!)S r (SRSS combination) (Algebraic combination) fP

• Frequency corresponding to the peak of the response spectrum Frequency higher than f at which the spectral acce,eration 1s well below the peak of the response spectrum Table B-1 gives estimated fP and fh values for the earthquake used in the study.

As explained below, ~ in general fs not known a priori since Sas given by Eq. B-86 is not knoim. This requires an iterative procedure to find the ultimate spectral acceleration capacity. However, generally~ ts very close to 1.0 and for practical purposes~ can be assumed to be 1.0. Effective Damping Ratio As an approxiaat1on the same expression for sliding damping (i.e., Eq. 8-48} was used for the rocking model, t.e.: (B-91) where: (B-92) (B-93) and where*b is a calibration tem as discussed below (i.e., see section

  *Determination of Effective Frequency and Damping*).

The Sillle form was chosen as for the sliding model since ~his multiplied by an empirical constant. It turned out that the same value of b was found for both models, that justified not developing an exact hysteretic loop for the rocking B-39

DRAFT model. Note that Eq. B-93 overstates the energy loss, which is reduced by the empirical constant b based on results from time history nonlinear analyses. ouct111tv scale Factor The ductility scale factor, F,, is the ratio between the earthquake level that causes failure and the earthquake level that causes yield. The ratio can be defined between peak ground acceleration values or between corresponding elastic response spectral ordinates (i.e., at the same frequency and damping~tio). The reference yield level fs defined in this report as the earthquake level that causes the fixed-base model (i.e., no base displacement} to reach the yield D!..1:9! in the weld spring. At the yield level the inertial force in the top mass times yh/w exactly equals the sum of the weld spring force plus the force due to gravity. This definition is contrasted to a more common definition that is sometimes used where the yield level is defined as the earthquake level that causes the weld spring to reach the yield displacement. The definition used here is preferred for the following two reasons:

• Structural analysts usually do not include the flexibility of welds in calculati~g a component frequency. Thus, defining the yield capacity at the fixed-base frequency is consistent with design practice.
• The yield displacement is negligibly small (e.g., 0.001 inch for a 3/16-inch weld size); therefore, the difference between yield capacities based on the two definitions is small.

The ductility scale factor can be expressed 1n terms of a reference earthquake level characterized by a family of elastic response spectra at different damping values. (B-94) where: Factor to scale the reference spectra to the level corresponding to failure (1.e., displacement 1n weld equal to the ultimate 111111 t) FSY

• Factor to scale the reference spectra to the level corresponding to yield (i.e., uplift B-40

DRAFT force at corner of the fixed-base model plus the force due to gravity} The FSu factor 1s just equal to the ultimate spectral acceleration capacity Sae, divided by the reference spectral acceleration at the effective frequency and effective.damping, Sa(fe, P.). Frona Eq. B-84 the spectral acceleration capacity sa. can be expressed in terms of the IIOdel properties. Noting that FmfW is equal to (0.5 + Cu)/1\i from Eqs. 8-61 and B-62, FSu becoaes: {B-95) The expression for FSY can be derived in a s1~11ar ~anner as FSu starting with a free body diagram. An alternate way to obtain FSv 1s to note for the fixed-base IIIOd,el with a frequency f 1 and damping P1 that there is no frequency shift. Hence, X and X1 b;oth equal to 1, and FME is just equal to O from Eq. 8-80. Hence using Eq. B-95: O.S+Cu g l (B-96) Sa(f1 , Pt) ~ (1 - a)e where Ry is equal to 1.0 since Eq. B-90 with FIil equal to 0 gives this result. Finally, combining Eqs. B-94,' B-95 and B-96: FJ

• X. Sa(f11 fltl l I (B-97)

X Sa(f., Pe) 1 + FME I\, Determ1nat1on of Effective Frequency and Damping In developing the pseudo linear-elastic model as presented above the concept of an effective frequency, f., arid effective damping, P., are introduced. Since the model frequency and damping change as the weld bec011es nonlinear, it is necessary to use effective or *average* values that are in between the fixed-base and secant model properties. It 1s through these two variables where the model is calibrated to fit the results of the 1nel~stic DRAIN-2D analyses. To this end the following eapirical relationships are defined: a (8-98) 1 - X... (1 - X) B-41

DRAFT' (B-99) where Ph is the maximum possible hysteretic damping at the secant frequency (i.e., maximum displacement) given by Eq. B-91. As discussed in Section 4 of the report, best fit values for a and b were found to be 1.6 and 0.3, respectively. Procedure for CaJcuJat1ng ouct1J1ty scale Factor The ductility scale factor, Fµ, is calculated following the equations given above. For the case where the weld yield force is known, iteration is required due to the term~ (see Eq. B-62), which is the ratio of the total base force, F,c* to the dynamic mode base force, F10 (see Figure B-10). This is because Eq. 8-90 for~ depends on the ratio S (see Eq. B-86) that in turn depends on the spectral acceleration capacity at the effective frequency and damping, Sae. The latter term is unknown a priori. ~ converges very quickly and for practical situations can be assumed to be 1.0 without a significant loss of accuracy. For this assumption no iteration is required. The more general problem is the determination of a reduced response spectrum Sar as discussed in Section 2 of this report. For this case both~ and Sar are unknown and two levels of iteration are required. However, in practical problems the two iterations are combined into a single process where both parameters are corrected simultaneously in each iteration. The steps for calculating FP using the simplified methodology for the rocking model are given in Tables 8-3 and B-5. Note in Table B-5 FHE is just equal to -FN (see Eq. B-80). Here the equation for Cu, the weld yield capacity coefficient, is written 1n terms of the reduced spectral acceleration, Sar. Using Eq. B-96 where FSY is equal to F&N (i.e., required minimum safety margin) and Sa(ff, Pt) is equal to Sar, (0.5 + Cu) is found directly to be: (0.5 + Cu)

• F111 s8r (I - <<) e (B-100) g For cases where the yield capacity, FY, for the weld is known, then Cu is just:

Cu

• F/W (B-101)

Figure B-7 shows a flow chart for calculating the response spectrU11 reduction factor, that is just equal to 1/F, (see Section 2). B-42

Table 8-5 SIMPLIFIED METHODOLOGY FOR ROCKING MODEL SteD Eauation Constant Mgm1liied Eff~ti~~ Er~gugnt~ (2 ff)2 ~ (Ce + fy)/g Kf Cocnponet fixed-base a 1t I. Basic Model Parameter, A A stiffness (0. 5 + Cu)/Ru H

• Component mass ll
• Fraction mass at base where: ff - I~

21t (1--u)H 6u ~ Weld ultimate displacement capacity Ce m _e2 (l-11) Th g

• Acceleration due to gravity e a

                                                             -w

• See note 1 tD ~

....w

                                                       - 0.25          N\,

I f1 + ~ h

• C0111ponent height wM Say. , w c Component width (0.5 + Cu) C FSM - (1 - <<) e g y
• Model he1ght/cOIIJ>Oflent height

2. Nonia.lized Secant Frequency Squared, X X - (1 +

2 cosO)( 1 + A) AMO C Incremental 11\ilSS IIIOIM!nt of inertia where 8 - sin-1 .fie a

• Empirical frequency constant (diaens1onless). A value of 1.6 was found to be appro-priate (see Section 4)

{ Ce f~ fil A B - (1 + A)2 F94 = Required minimum safety factor

Table B-5 (continued) Step Equation Constant

3. Normalized Effective Frequency, Xe \ - I - (l-X) 1 Eff~,ti~~ DM!l!i!l9 B~tio

4. Elastic 08lllp1ng Ratio at ,fp . ~ rx pf a Elastic dUIJ)ing ratio at Effective Ultimate Drift, fixed-base frequency P1p b

• Empirical daaptng a,

S. Hysteretic Weld Damping Ratio at Ultimate Ortft, ~ ~ "' b ~ (1 -i) [1 - ~] constant (dimensionless). A value of 0.3 was found to I .,., be appropriate (see Section 4) where D ., ~ I { ~] (I - X)2 I\, -

                                                                                 &y 6y

• Weld yield displaceaent

6. Effective Duiping Ratio at Effective Frequency, Pe Pe - -

X Xe (P,p + [\j)

Table B-5 (continued) Step Equation Constant 0Y~t1Jjtr s,11~ Eittgr

7. Fraction of Total Ff.£ =

                                                  +[ 2]  X (1-X)

Mass Excess to D Dynamic Mode, Ftf: C0 I U1

8. Ductility Scale Factor, F 11 FIi D (x. ]['*(ff, X Sa(fe*

Pf)]( I fie) 1+FHE

                                                                          ]( I
                                                                            *Ru l   Sa(f. P)

• spectral acceleration at frequency, f, and damping, P where: fe
• ff {x; R.. detenained from Table B-3 using f = f and Pd P (Note that R.. is assumed initially for the first trial and

       ~vised values are obtained using Table\-3 for th subsequent 1rerat1ons. For rocking cases, ~ 1s close to unity and Ff£ is close to 0)

Example calculation Figure 8-81 shows an ex1J1Ple 5-percent damped unifona hazard spectrum for a EUS site. Figure 8-Sb shows the corresponding reference spectrum for different damping values. For this exllll)le the reduced 5-percent dutped spectral ordinate, Sar, ~t the fixed-base frequency of 15 Hz is calculated for three trials as tabulated in Table 8-6. Note that the model p1ra1119ters assUlled in the analysis are provided at the bott011 of the table. Initially in trial 1, values of 0.250g and 1.0 were assUll&d for Sar and f\i, respectiveiy. The calculation steps in the flow chart shown in Figure 8-7 and the equations given in Tables 8-3 and 8-5 were followed. At the end of each trial new values of Sar and Ru (however, Ru was found to be essentially 1.0) were calculated that were used in the subsequent trial. By the third trial convergence was obtained. Thus, the evaluation spectrum is reduced from Sa 1 equal to 0.507g to a value of Sar equal to 0.271g. REFERENCES

1. N. H. Newmark and E. Rosenblueth. Fundamentals of Earthquake Engineering.

Prentice-Hall, Inc., 1971. 8-46

Table B-6 EXAMPLE ROCKING MODEL ANALYSIS Trial ~ I\, CU A B 8 X \ (fl) D ~ ~ -1_ _k_ Sa(~ ~It !at_ 5fgf 0.250 ""f.o 0.438 2.125 0.017 0.194 0.323 0.464 10.217 1.018 0.028 0.117 0.101 0.009 0.381 1.895 0.268 2 0.268 1.0 0.505 l.!113 0.017 0.lffi 0.338 0.483 10.428 1.018 0.029 0.115 0.101 0.009 0.383 1.875 0.270 3 0.270 1.0 0.515 1.963 0.017 0.187 0.340 0.48> 10.460 1.017 0.029 0.114 0.101 0.009 0.384 1.872 0.271 m I lbtel Para,et.ers (Pennssible inelastic ancror displacarmt .. 0.01 1rdt)

                  * .. 0.25     Sai .. Sa(ff'  't>  - o.rm g      fa25Hz p                    ~
                                                                                             . 8.333 ff

• 15 Hz i
• 0.46 g fha33Hz e - 3.333

                 ~f

• O.ffi 11w

                                                    ., 10          h .. 100 inches      fl   a 0.333 F!H .. 1.5                                         W   a 20 inches
                 &u

• 0.01 irdt

DilAFT Appendix C EXAMPLE BUILDING RESPONSE AMPLIFICATION

Appendix C EXAMPLE BUILDING RESPONSE AMPLI~ICATION This appendix presents an example analysis to demonstrate that the expected amplification for in-structure spectra for typical nuclear power plant structures realistically should not exceed 2.0 for frequencies greater than about 10 Hz for structures with fundamental frequencies less than about 10 Hz. This conclusion assumes that the ground response spectrum is falling at lower frequencies {i.e., 2 to 10 Hz), that is typical of UHS developed for the eastern U.S. {EUS). However, for the purposes of reducing ground response spectra in the high-frequency region, this is not a significant constraint as discussed below. Experience indicates that fundamental frequencies of massive nuclear power plant structures are less than 10 Hz, even at rock sites. Although, soae past design . analyses have computed higher frequencies. It is believed that if realistic modeling assumptions are made for foundation stiffnesses fundamental frequencies would not exceed 10 Hz. An example shear bea111 building consisting of 6 equal lumped masses and equal shear elements was developed and is shown in Figure C-1. The properties 'of the 110del were selected to achieve two goals:

• Fundamental frequency is less than 10 Hz
• The second and third frequency are in the power of the earthquake input {i.e., 15 to 30 Hz)

By using shear stiffness elements this was achieved as shown in Figure C-1 for the target ground response spectrum that ts shown in Figure C-2. Raleigh damping was used tn the analysis, and 7 percent damping was assigned to the first and second modes. The 1110dal duaping values for all modes are also shown in Figure C-1. A synthetic time history was develop~ whose response spectrum closely atches t~e target response spectrU11 as shown in Figure C-2. The time history was input to the model in Figure C-1 and the in-structure response spectrum for the top 111ss (that is the maximum poin~ of response) was calculated. The in-structure and C-2

ORAfl Frequency Damping Mode (Hz) p 1 6.0 0.07 2 17.M 0.07 3 28.26 0.09 4 37.26 0.12 6 -44.10 0.14 6 -48.36 0.15 Figure C-1. Shear Beam Building Hodel ground response spectra are plotted in Figure C-3. The ratios of the ordinates of the two response spectra shown in Figure C-3 can be visually estimated. For frequencies between 10 and 30 Hz the calculated maximum ratio of the in-structure response spectrum to the ground response spectrum was found to be *1.89. At 50 Hz, which is close to the peak ground acceleration (PGA), the ratio is 1.60. Note that the PGA is controlled primarily by the ground response spectral ordinate at the fundamental frequency of the model (i.e., 6.0 Hz). Based on the recOll'mlendations in Section 2 for reducing the ground response spectrum no reduction will occur below 8 Hz. Hence, the PGA of the floor response spectrum will not be significantly affected by the reduction procedure. It is concluded that the shape of the ground response spectrum below 10 Hz is not critical to the use of an amplification of 2.0 for developing the response spectrum reduction factor for frequencies greater than 10 Hz. In other words the PGA of in-structure C-3

CTl z-0

           ~

n::: w _J w u 0.1 u

           <(

_J

           <(

n::: 1-- u w 0... {/) 0.01 0.1 1 10 100 FREQUENCY, hz Figure C-2. Target and Artificial Time History Ground Response Spectra z 0

           ~

n::: w _J w u u<( _J

            ~ 0.1 1--

u w 0... (/) 0.01 0.1 1 10 100 FREQUENCY, hz Figure C-3. Comparison of Response Spectra at the Top Hass and Ground Levels C-4

DAAR response spectra should be about the same with or without any reduction taken above 10 Hz. This exutple is not meant to be the most limiting case that could be theoretically derived. It is intended to represent a reasonably conservative case to indicate the nature of the issue and the likely upper bound results that would be obtained for practical problems. The use of a amplification factor of 2.0 is realistic and adequate for developing the reducing factor to apply to ground response spectra in the high frequency region that will be used to generate in-structure response spectra. C-5

Appendix D EARTHQUIKE RECORDS USED IN THE STUDY

Appendix D EARTHQUAKE RECORDS USED IN THE STUDY Table 0-1 lists the 15 earthquake records which were used in the high-frequency study. Record 1 is an artificial time history that was fit to the NRC Regulatory Guide 1.60 horizontal ground response spectrum. Record 12 (Leroy Modified) 1s the same as Record 13 (Leroy, Ohio} except the t1me step has been increased by the factor 1.25, which causes the corresponding response spectrum to peak near 18 Hz {as compared to 23 Hz for Record 13). Record 15 is an artificial time history with significant high-frequency content. The time history and response spectra plots are shown on Figures D-1 through D-15. Note that all records have been scaled such that the peak 5 percent-damped spectral accelerations are all equal to 0.8 g. 0-2

Table D-1 EARTHQUAKE RECORDS USED IN HIGH-FREQUENCY STUDY Station Magnitude Site Intensity lmi. Eirtbi;iuak~ Dit!l ltiU t!lll!l!mumt 'Ms MMI l R.G. 1.60 (Artificial) 2 01,Yll'Pia, WA 04/13/1949 Highway -Test Labs H86E 7.0 VIII 3 Parkfteld, CA 06/27/1966 Cholame No. 2 N65E 6.4 VII 4 Tabas, Iran 09/16/1978 Tabas Trans. 7.7 X 5 llll)erial Valley, CA 10/15/1979 E.C. Array No. 5 140 6.9 VII 6 Nahanni, Canada 12/23/1985 Site 1 long. 6.9 IX 0 w I 7 Saguenay, Canada 11/25/1988 Site 20 Long. 6.0 8 Gazl1, USSR 05/17/1976 Karakyr Point East 7.0 IX 9 Bear Valley, CA 09/04/1972 Kelendy Ranch N29W 4.3 VI 10 Gazli, USSR 05/17/1976 Karakyr Point North 7.0 IX 11 Saguenay, Canada 11/25/1988 Site 16 Long. 6.0 12 Leroy Modified 13 Leroy, Ohio 01/31/1986 Perry NPP Basetiat South 4.8* y 14 New Brunswick, 03/31/1982 Mitchell Lake 28 4.0* IV Canada 15 Artificial C,

  "s

• Surface wave magnitude
• Moflent magnitude

0.300 (_') 0 225 KA* O.Z4Z I z- 0.150 O 0075

     ~

0:::: 0.000 w _j -0,075 w U -0 150 u

      <( -0.225
            -0.300 _ J _ _ . - ~ ~ ~ ~ ~ - . - . -.........,........,,-,-,-rT.,...,.""T"T"TT"-r-r-,-,-,-r-r-,-,-T"T"TT",-,-,-,-,r-,

0.00 5.00 10 00 15.00 20.00 25.00 TIME, SEC.

a. Time History

1. 00 -,--,--T"""T-rTTTT-r-----.---,--,--.-,-rrr,----,----,--.---,....,....~

CJ) 0.90 z 4\, S\, 6\, 8\, 10,. 12\, 15\, Q0.80 18\, 21\, ZS\ 0-INO f-(2 0.70 w _j 0.60 w u o.so 0

                     <(

0.40 _J

                     <(

O:::: 0.30 f-u W0.20 0... (/)0.10 0, 0 0 +--i--T"""T-r-rTTT-r-----.---,--,--.-,-rrr,,-----,----,---r-....,....~

0. 1 10 100 FREQUENCY, hz

b. Response Spectra Figure D-1. Scaled R.G. 1.60 (Artificial) Earthquake Time History and Response Spectra D-4

DRAFT 0.300 0 0.225 PIA

• 0.:104 I z 0.150 O 0.075

    ~

0:::- 0.000 w _J -0.075 w U-0.150 u

    <C -0.225
         -0 .300 -t-,..,..,.........,..,.. .-+,..,..,...,..,....,.,"T"T"T"T.,...,."T"T.,...,.......,."'T'"T-r-1".......,............,...,....,...,.~~

20.00 0.00 40.00 60.00 80.00

                                                                   .TIME, SEC.

a. Time History 1.QO -r--r-.-r-..-rTTMc-----,---,-...,.............,~-~--~

CJl 0.90 z "* 10\, "* 12111,"*16\, 8\, Q0.80 111, zn. zn r-: D-JNA

               ~0.70 w

_J 0.60 w* u o.so 0

               <(

0.40

               ~*

o::: 0.30 1--- uwo.20 0.. (I) 0.10 0.1 10 100 FREQUENCY, hz

b. Response Spectra Figure 0-2. Scaled Olympiat WA Earthquake Time History and Response Spectra D-5

0.300 (.9 0.225 l'M

• 0,241 O z- 0.150 0 0.075

       ~

n::: 0.000 w _j -0.075 w U -0.150 u<( -0.225

           -0.300 0.00               10.00                 20.00                          30.00                          40.00 TIME, SEC.

a. Time History 1.00 -,--.--r-r.,...,-.rrTT---,-....,.....,....,.-.-.-,...,...-..-..--,-...,....,.....,..,,

CJ) 0.90 z "'* 5't, 10\, 12\,'"*1S\, 8\, Q0.80 18\, 21,, 25\ DAMPI~ l-

                ~0.70 w

_j 0.60 w 8o.5o

                <(

0.40 _J

                ~0.30 l-t3   0.20 (L

(I) 0.10 0.00 _ __,,_.,,.....,...,...,...,"TTT---,--,-.,......,......,..,.......--,--.......,...........rrr1 0.1 1 10 100 FREQUENCY, hz

b. Response Spectra Figure 0-3. Scaled Parkfield, CA Earthquake Time History and Response Spectra D-6

0.300 (.'.) 0.225 NA* O.lff t

         ~

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0.00 5.00 10.00 15.00 20.00 25.00 30.00 TIME, SEC.

a. Time History 1.00 --.--,-...--,-.,..............,-__,..__,........,.~~-~--~

CJ\ 0.90 z 4111, 5\, "* 8\, 10\, lZ\, 15', Q0.80 18\, 21\, 25' 0-IIIG I-

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u 00.50

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                     ~0.30 1-u wo.20 (l_

(I) 0.10 0.00 - - -...............-........ 0.1 (

                                                                                 ----~-----........!                   10                                               100 FREQUl;NCY, hz

b. Response Spectra Figure D-4. Scaled Tabas, Iran Earthquake Time History and Response Spectra 0-7

0 300 C) 0.225 1"11

• O.llJ 9 z- 0.150 O 0.075

  ~ 0 000 et::

w _J -0.075 w U -0.150 u

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0 00 10.00 20.00 30 00 40.00 TIME, SEC.

a. Time History 1.00 -r--r--T""T"TTTTT..----r----.---,-..-r-r-rn----r---r--.-,.......,..T"M 01 0.90 4,, s,, &,. a,,

z 10,, 12,, 1s,, Q0.80 1e,, zn, 2s, DAMPINII I--

               ~ 0.70 w

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0. 0 0 +---r--T"""T--r-r"TTT..-------.--,.............-.-.....-,,----,---,---,--,..........,....l 0.1 1 10 100 FREQUE;_NCY, hz

b. Response Spectra Figure 0-5. Scaled Imperial Valley, CA (E.C. Array No. 5} Earthquake Time History and Response Spectra 0-8

0.300 C., 0.225 PM* O.Ul I

z. 0.150 O 0.075

 ~ 0.000 ....."'111111111 0:::

w _J -0.075 w U -0.150 u

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0.00 5.00 10.00 15.00 20.00 25.00 TIME, SEC.

a. Time History 1.00 .----r-,......,..-.rrn,----,----.-"T""T"..,...,...,...,...._~~--

CJ) 0.90 z "* n. 6\, II\, 10\, 12\, 15\, Q0.80 18%, 21\, 25' 0-INll I-

             ~ 0.70 w

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             <(

0.40 _J

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              ~0.20 0....

U) 0.10 0.1 1 10 100 FREQUEN_CY, hz

b. Response Spectra Figure 0*6. Scaled Nahanni, Canada Earthquake Time History and.Response Spectra

0.300 C) 0.225 PIA* O,lll I z- 0.150 O 0.075

    ~ 0.000 0:::

w 0.075 w U -0.150 u

    <( -0.225
         -0.300 ...............,...,..,~..........,...,....,....,...,....,........,....,......+,,..,......,.....,.........,...,..,,...,...,....,..................,..,,...,...,....,........,...,....,..,...,~

0.00 5.00 15.00 20.00 25.00 SEC.

a. Time History 1.00 --,---,-,,-,-.,......,~--.--.-.,...,....,....,..........--,.........,,...........,.......,.......,

CJ) 0.90 4,, 5'", 6,, R, z Q0.80 10%, 1n, 1n, 18', 21\, ZS\ 0-1110 1-

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b. Response Spectra Figure D-7. Scaled Saguenay, Canada (Site 20) Earthquake Time History and Response Spectra D-10

DRAFT 0.300 l'IIA

• o.m 1 C., 0.225 z 0.150 O 0.075

    ~ 0.000 -+-~~Ml et:

w _j-0.075 w U-0.150 u

    <( -0.225 2.00             4.00         6.00              8.00               10.00     12.00   14.00 TIME, SEC.

a. Time History 1.00 --,--,--,...,....,..,..TTr,---~-.-..,.....,....,.........,..,....._ _~-~

Ol 0.90 z Q0.80 ""* 10\, n, 12',"*15\,"* lM, 21', ZS\ 0-INII l-

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b. Response Spectra Figure D-8. Scaled Gazli, USSR (Karakyr Point - East) Earthquake Time History and Response Spectra D-11

0.300 (.'.) 0.225 z 0.150

                                                                                   ' -o.w' O 0.075
   ~

0:::: 0.000 w __.J -0.075 w U -0.150 u<( -0 225

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0.00 5.00 10.00 15.00 20.00 25.00 TIME, SEC.

a. Time History 1 .00 _ _ _.....,.....,...,..,...,...,----,----,-..................,..,.,,_----~

01 0.90 zQ0.80 C\, SI, 6\, 81, 10\, 12,. 151i, lM, Zll, 25\ l- DAMPING

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b. Response Spectra Figure 0-9. Scaled Bear Valley, CA Earthquake Time History and Response Spectra 0-12

DRAFT 0.300 0 0.225 PIA" 0.171 I 6.00 8.00 10.00 12.00 14.00 TIME, SEC.

a. Time History 1.00 ....--,--r-r..,...,.....,..,.,-~...................,....,...,M"T'""--,----,--,---,-,-n-n C]l 0.90 41, 5', 6\, II\,

z IOI, 1n, 1n, l.M, Zl'I, ZS\ Q0.80 0-ua r

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           ....J 0.60 w

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(/) 0.10 0.00 4----.,....,.;F,-..,...,..n-rr--,---.-...........,...,.,.,.....-,--,--r-T"T"'l"Tm 0.1 1 10 100 FREQUENCY, hz

b. Response Spectra Figure D-10. Scaled Gazlf, USSR (Karakyr Point -_North) Earthquake Time' History and Response Spectra D-13

0.300 (5 0.225 PIA* O,W t z- 0.150 O 0.075

  ~

0::: 0.000 w

  ....J -0.075 w

U-0.150 u

  <( -0 225
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0.00 10.00 20.00 .30.00 40.00 TIME, SEC.

a. Time History 1.00 -.--..--,---,-..,....,..,.....,...---,--,-......,.........,..,...--..-..........,...,...........,

CJ"l 0.90 4%, n, 6%, A, z Q0.80 10%, lZ\, 15\, 1s,. tn, tn DAMPZIIG I--

              ~ 0.70 w
              .....! 0.60 w

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b. Response Spectra Figure D-11. Scaled Saguenay, Canada (Site 16) Earthquake Time History and Response Spectra 0-14

DRAFT 0.300 ('.) 0.225 PGA

• 0.176 V

        -0.300 -+-,-..,....,....--,-,...,....,'"T'"'t'..,..,....,....,...,...,....,,--,-,-.-r-r-..,...,......,...,,..,...,....,...,.......,.......,....,....,_,._,...,.......,...,....,....'"T'"'t'......,....,...,

0.00 5.00 10.00 15.00 20.00 25.00 TIME, SEC.

a. Time History 1.00 -.---.---r--r-..,..,...,..,..,.,--,--,----r-"T""T......,..,r-n---.---.---r.....-r'TTTI CJ) 0.90 z 41i, !iii, "

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b. Response Spectra Figure D-12. Scaled Leroy Modified Earthquake Time History and Response Spectra 0-15

0.300 0 0.225 NA

• 0.171 t
• -0.300 -h-rT-,-.,,..T"T""l""T""'li""T"'"T""T"T"T"T""T'"'r",...,...,....,...,,...,...,....,..,....,....,....,...,....,..,C""T""T...,...........,.......,.._...,........,

0.00 5.00 10.00 15.00 20.00 25.00 TIME, SEC.

a. Time History 1.00 ,--,--.,......,---r-r-rr,-.,.....--,.--,.--r-.,..,....,...........--.....--.-..,-,-~

0) 0.90 z 4\, 5', 6'11, Ills, 10,, 12'11, 1S\, 0 0.80 111\, Zl\, Z5' D-INII I-

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0. 1 10 100 FREQUENCY, hz

b. Response Spectra Figure D-13. Scaled Leroy, Ohio (Perry NPP Basemat) Earthquake Time History and Response Spectra 0-16

0.300 (.'.) 0.225* z 0.150 PIIA

• O,UO I O 0.075

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et:: 0.000 w

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U-0.150 u

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         -0.300   -..~-~~-~~~~~---~---~----

0.00 -2.00 4 0 6.00 8.00 10.00 TIME, SEC.

a. Time History 1.00 --r--.-,--,-..,...,...,..,..,......----,..---,.....,....~~--~~~

O"I 0.90 z "* 10\, 5111, 11'1, 12\,"*15111, Q0.80 18\, Zl\, ZS\

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(.I) 0.10 0.1 1 10 100 FREQUENCY, hz

b. Response Spectra Figure 0-14. Scaled New Brunswick, Canada (Mitchell Lake) Earthquake Time History and Response Spectra 0-17

0.300 (..'.) 0.225 z 0.150 O 0.075

     ~

n::: 0.000 w _J -,0.075 w U-0.150 u

     <( -0.225

5. 0 10.00 15.00 20.00 TIME, SEC.

a. T1me History

1. 00 -,--.----,........--,-,--,..,.,...---,---,-..............,,.....---,--.,.......,--,-,..,.,..,.,

CJ) A 0.90 4\, H, 6\, 8%, z 10\, 12', 151, Q0.80 18', zn. zn 0-IIIICI f-

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b. Response Spectra Figure 0-15. Scaled Artificial Earthquake Time History and Response Spectra 0-18

,/ BADAN TENAGA ATOM NASIONAL ( NATIONAL ATOMIC ENERGY AGENCY) ~ Jl. K.H. Abdul Rohim, Kuningan Barat, Marnpang Prapatan ~ P.O. Box 86 Kby. Jakarta 12710 Indonesia Phone : 611109, Telex : 62364, Cable : Batan Jakarta

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Secretary, U.S. Nuclear Regulatory Commission - .- ..,.,

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                                                                                              -<      C Washington, DC 20555,                                                       . *"       )

N ("' U.S.A. Att  : Docketing and Service Brand I

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Dear Sir,

                                                                               -- '.l Ul After reviewing the US - NRC proposed changes of rules of reactor site criteria for future nuclear power plants,                        published in Federal Register vol. 57 No. 203, dated October 20,                      1992, page 47802 through 47821, we would l i ke to make comments as follows :

1. In our knowledge the proposed change to decouple reactor siting from reactor design is contrary to the internationally-accepted practice of consider i ng the engineered safety features in making reactor site decision. However, we could understand the reasons for the proposed change as far as the risk of the nuclear reactor to the public and environment i s very low. In this regard further improvements of the safety in reactor designs should be conducted cont i nuously to gain a more safe and reliable reactor design.

2 . The numerical criteria for population density as well as distance to exclusion boundary in our opinion are irre l evant to safety goal, since the philosohy underlining the safety goal is the restriction of a risk arising from operation of a nuclear power plant, in comparison with other r'sks to which the public is exposed. On the other hand, the minimum distance to the exclusion area boundary of 0.4 miles (640 meters) will cause some problems to nuclear power plants having power level l ess than 3800 Megawatts, especially in finding a suitable site for the plants. Based on those reasons, we prefer to provide the numerical values of population density in a regulatory guide and the exclusion area distance to be allowed to vary according to power level .

3. For the proposed rule concerning the use of both the deterministic and the probabilistic se i smic analysis approach in determ i ning a safe shutdown earthquake to allow more

,I -.. NATIONAL ATOMIC ENERGY AGENCY JAKARTA - INDONESIA informed judgement, we could understand the reasons for this. However, it is not yet clear i f probabilistic seismic analysis methodology and its application in nuclear reactor regulation are matured enough or not, since the deterministic approach has worked reasonably well for the past two decades, in the sense that SSEs for plants sited wi th this approach are judged to be suitably conservative. Thank you for your attention.

l?' AC S :t: ~l :X L E NATIONAL ATOMIC BNBRGY AG8NC1 (13J\DI\N T8NAGA ATON NASIONAL) JAKA~TA - INUON~SIA NO 485/Fax/ DJ DATE 26 May 1993 NUMBER OF P.~GES SENT INCLUDING COVER SHEET 3 Secreta r y TO U. S . Nuclear Regulatory Commission FAX NR. : 1 301 5043509 Washington DC 20555, USA Att. : Docketing and Service Brand ffiCM Djali Ah im sa FA.'\ NR.:

• 1021) 511 110 Director General

- MESSAGE Please find       attached comments on proposed rev ision to            CF R   100 from BATAN.

Yours sincerely, Djali lb Director Gene ra l For tr:"-~smission difficulties contact

                                                                *93 HAY 24 A11 :03 May 17, 1993 U.S. Nuclear Regulatory Commission Docketing and Service Branch Washington D.C. 20555

Dear Mr. Secretary:

Summarized below are comments related to the U.S. Nuclear Regulatory Commission's (NRC) Regulation Appendix B to 10 CFR Part 100, "Criteria for the Seismic and Geologic Siting of Nuclear Power Plants, and associated supporting Regulatory Guide, DG-1015, "Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motion." The NRC should be commended for taking the step of revising the seismic and geologic siting criteria, considering the difficulty of the technical issues involved and the wide diversity of views regarding how the regulation should be formulated. While the regulation and supporting regulatory guides incorporate technical advancements that have been made over the past two decades, the overall objective of providing for a stable regulatory basis for seismic and geologic siting of future nuclear powerplants has not been achieved. In contrast, our review indicates that the regulatory implementation of the approach proposed may be more difficult than that which exists in the current regulation. The primary reason for reaching this judgement relates to the approach proposed for completing parallel deterministic and probabilistic analyses. The regulatory requirements pertaining to the Determination of Deterministic Source Earthquakes are likely to result in excessive technical and regulatory debate regarding the definition of seismic sources and deterministic source earthquakes. This is particularly true in the Eastern United States where wide diversity exists in the opinions of experts regarding seismic sources and earthquake occurrence rates. Past insights gained from licensing of nuclear powerplants do not appear to have been explicitly considered in developing the revised regulation. These issues include the use of the Central Stable Region as a single seismic source, the occurrence of an earthquake larger than the historic maximum for a given source (such as the New Brunswick earthquake), and the definition of the seismic source for both the New Madrid region and the Charleston region. These issues have resulted in regulatory delays for applicants. JUL 3 t 1993 Acknowledged by card-***"*--***........... __ _

    -~ ** ... , *     *** *
    *     *    ,.. I i ' *-..:.*;* ,,J :*

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2 It is not clear how the Deterministic Source Earthquakes will be defined given the range of seismic source information provided in the Lawrence Livermore National Laboratory (LLNL) and Electric Power Research Institute (EPRI) seismic hazard studies. The range of potential use (and miss-use) of the LLNL and EPRI information may result in excessive regulatory debate and delay. The enclosure to this letter outlines an alternative approach to the assessment of seismic and geologic design bases. This approach provides for the integrated use of deterministic and probabilistic methods to select the Safe Shutdown Earthquake Ground Motion. Such an approach allows for the strengths of each method to be used, and could result in a more stable regulatory framework. Such an integrated approach is compatible with the use of the LLNL and EPRI information without over reliance on the bottom line probabilistic numbers specific to each project. We have included in the

• enclosure several examples of how the approach could be implemented .

We thank the NRC for considering our comments, and welcome any opportunity to discuss issues and concepts with the NRC s.taff. Sincerely,

                                       ~            imball, Director Facilities Engineering Division Office of Engineering and Operations Support Defense Programs Enclosures cc:

Dr. A. Murphy, NRC

ENCLOSURE OUTLINE OF PROPOSED APPROACH FOR DETERMINATION OF THE SAFE SHUTDOWN EARTHQUAKE IN THE EASTERN UNITED STATES Provided below is an outline of a proposed approach for determining the Safe Shutdown Earthquake (SSE) in the Eastern United States (EUS) which integrates deterministic and probabilistic information. The discussion is linked to information provided in Draft Regulatory Guide, DG-1015, "Identification and Characterization of Seismic Sources, Deterministic Source Earthquakes, and Ground Motion.ft The overall proposed approach is based on the concept that three steps are necessary to develop the SSE. These steps are: Geological, Seismological, and Geotechnical Investigation (Step I); Identification of Seismic Sources and Magnitudes (Step II); Characterization of Ground Motion (Step III). The results of these steps is the definition of SSE Ground Motion (Step IV). Each of these steps is discussed below including comparison to information in DG-1015. Step I: Geological, Seismological and Geotechnical Investigations It is proposed, similar to DG-1015, that Geological, Seismological and Geotechnical (GSG) investigations be used to identify and characterize seismic sources. Appendix D to DG-1015 describes adequate investigation procedures to complete this step. In the context of the EUS, results of this step would be compared to the range of seismic source information contained in the Lawrence Livermore National Laboratory (LLNL) and Electric Power Research Institute (EPRI) studies (Step 2). As such it is recommended that DG-1015 be revised to explicitly state that the investigations within 40 km (25 miles) and 8 km (5 miles) be reviewed to determine if an active seismogenic source exists in close proximity to the proposed site that has not been considered in the LLNL and EPRI results. If this were the case revised probabilistic seismic hazard calculations would be necessary

• Step I is also necessary to ensure that site specific geotechnical factors are assessed including the potential for ground motion modification through soil.

Step II: Identification of Seismic Sources and Magnitudes Step II of the proposed process would eliminate the Deterministic Source Earthquake (DSES) and revise the Probabilistic Seismic Hazard Analysis and Controlling Earthquake discussion in DG-1015. The primary reason for this reco111nendation relates to an incomplete understanding of how LLNL and EPRI information would be used to define seismic sources and DSES in the EUS. Given the seismic source maps (and alternatives) and earthquake occurrence parameters (including upper magnitudes), from the experts and expert teams, it is unclear how a stable regulatory framework could be implemented. The U.S. Nuclear Regulatory Commission (NRC) and the applicant will be spending resources defending why the worst case from the experts is not used to define the DSES.

2 As an alternate proposal it is recommended that, in the EUS, the collection of seismic sources and occurrence rates be used to derive the magnitudes (M's) and distances (D's) which most contribute to the seismic hazard at the site. The primary advantage of this approach is that there is explicit recognition that the causative mechanisms of earthquakes in the EUS is uncertain, and that this uncertainty is considered in selecting Mand D. The steps to implement this approach are as follows:

1. A probabilistic Seismic Hazard Analysis (PSHA) should be completed at the site. Either LLNL or EPRI should be used in the EUS.

2. The PSHA should be deaggregated to define Mand D, at a minimum of 2 response frequencies. The first frequency should be greater than about 5 to 10 hertz and the second frequency should be less than 5 hertz. Similar to DG-1015 the probabilities could be linked to the 50th percentile of exceeding the current population of reactor safe shutdown earthquakes using either LLNL or EPRI data (DG-1015 Appendix B).

3. The results would be M1 D1 for the high frequency and M2 D2 for the low frequency. For each case Mand D would be the mean Mand D which contributes most at the selected probability.

4. Any historic earthquake greater than M = 6.25 should be used as an additional Mand D with the distance selected using the actual historic location. This will assure that ground motion from larger historic earthquakes is properly accounted for.

The result of the above steps are the inputs to deterministic ground motion calculations. Step III: Characterization of Ground Motion This step of the integrated process would be to develop 84th percentile estimates of vibratory ground motion for each Mand Dusing procedures such as described in Standard Review Plan Section 2.5.2, "Vibratory Ground Motion." This step of the process would be the same as has been used in past licensing of reactors. The applicant should then be given the choice of using the computed ground motion estimates or using a standard spectral shape to envelope the estimates. Examples are provided later in this write-up which show that the standard spectral shape provided by NUREG/CR-0098 (Newmark/Hall) is adequate for modeling the ground motions from the individual earthquakes for frequencies less than about 10 to 15 Hertz. Step IV: Develop Safe Shutdown Earthguake Ground Motjon The Safe Shutdown Earthquake {SSE) ground motion is the envelope of the individual ground motion estimates completed in #3 above. The above approach retains the deterministic characterization of a site and provides for deterministic estimates of vibratory ground motion while recognizing that probabilistic results provide the most useful information regarding potential seismic sources and resulting magnitudes and distances.

3 The above approach would eliminate the following issues which have required large staff resources and caused past licensing delays: o Use of the Central Stable Region and magnitude; o Northern and southern limit of the New Madrid seismic zone; o Need to define tectonic provinces such as the Michigan 6asin; o Debate over whether the Anna, Ohio earthquak~s are related to a specific seismic sources; o The cause of the Charleston earthquake; o The occurrence of an earthquake sltghtly larger than the historic maximum such as the New Brunswick, Canada earthquake;- o Definition of tectpnic provinces for New England; and o Definition of the minimum size of floating earthquake. Examples of Proposed Approach Provided below are several examples of implementing the proposed approach to computing ground motion estimates. Table 1 shows the seven sites used as examples, and the Mand D which most contribute to the seismic hazard at each site. The Mand D were provided by LLNL and are based on using the median probabilities of exceedance of the population of existing nuclear power plant sites as listed in DG-1015. Two-estimates of Mand Dare provided' for each site, one deaggregating the probabilistic results using the average of 1 Hertz (Hz) and 2.5 Hz and one deaggregating the probabilistic results using the average of 5 Hz and 10 Hz. Five of the examples sites are classified as rock sites while two of the sites are classified as deep soil sites. - The ground motion attenuation equation developed by McGuire, et al., (1988 EPRI NP-6075) was used to derive response spectra for the M, D pairs shown on Table 1. For the two deep soil sites the frequency dependent soil -(Category V from Table 6-5) correction factors from McGuire, et al. (1988) were also used. Figures 1 to 7 display the resulting response spectra compared to the 84th percentile response spectra from Newmark and Hall (NUREG-CR/0098) anchored to peak ground accelerations of 0.10g and 0.20g. The Newmark and Hall shapes at peak accelerations of 0.10g and 0.20g were selected as representing the range of values associated with the current population of reactors. Figures 1-A to 7-A were based on assigning 50 Hz as the frequency of the peak acceleration for the Newmark and Hall estimates. Figures 1-B to 7-B were based on assigning 15 Hz as the corner frequency for the constant acceleration portion of the Newmark and Hall estimates. The following observations are based on reviewing Figures 1-7.

4 o The Newmark and Hall estimates at peak accelerations of 0.10g to 0.20g are conservative for frequencies less than about 10 Hz for all sites, including Seabrook. o The Newmark and Hall estimates are not conservative for frequencies above about 15 Hz. o The Newmark and Hall spectral shape is not matched well in the EUS using the McGuire, et al. (1988) model. Based on the above the following recommendations are provided. o It is recommended that NRC select an attenuation model and calculate the 84th percentile spectral shape for each Mand D for all reactor sites. If the selected model provides for adequate results when compared to Newmark and Hall spectral shapes, then the model should be the preferred model listed in Section 2.5.2 of the Standard Review Plan and DG-1015. o It is recommended that the Newmark and Hall spectral shape be provided as the preferred standard spectral shape. The estimates provided using the ground motion attenuation model should be compared to both median and 84th percentile Newmark and Hall spectra. A determination should be made whether the corner frequencies in Newmark and Hall should be modified to higher frequencies more consistent with the EUS ground motion and whether the ratio of peak velocity to peak acceleration be modified to better represent the Mand D from the probabilistic results. o It is recommended that an evaluation be completed to determine the most appropriate ground motion value to scale the Newmark and Hall spectral shape in the ELIS. For example, it may be most appropriate to scale the Newmark and Hall spectral shape to the 10 Hz ground motion estimates. Based on Figures 1-7 it does not appear appropriate to scale to peak acceleration. o It is recommended that the NRC develop a position in a regulatory guide on the significance of ground motion estimates for frequencies above 15 Hz, with respect to how they should be treated in design of nuclear power plants.

TABLE 1

SUMMARY

OF DOMINANT MAGNITUDES AND DISTANCES Avg. 1/2.5 Hz Avg. 5/10 Hz Site Site Conditions M D M D Limerick 5.75 25 5.60 19.6 Rock Braidwood 5.95 103 5.50 23 Rock River Bend 6.4 219 5.35 25 Deep Soil Vogtle 6.3 110 5.75 40 Deep Soi 1 Seabrook 5.9 20.8 5.75 18.1 Rock Fenni 5.65 36 5.5 25 Rock Comanche Peak 5.70 95 5.3 28 Rock

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 1-A LIMERICK (ROCK) 8 1  :  : p E 0.9 + **** *H O 000 0 00

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RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 1-B LIMERICK (ROCK) s 1 p ..... , ... *: .... E 0.9 C T 0.8 R A 0.7 . . .. . . .. . ... .. L 0.6 . ....... . . A C C 0.5 .. ... .. .* ...... .:. .. E L 0.4 E R .. : . A 0.3 T I 0.2 0 N 0.1 ..  : ... .. . . g 0 1 10 100 FREQUENCY (HZ)

               - - NIH PGA*.1                 -+- NIH  PGA*.2   --3/4- M1 D1   ~        M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 2-A BRAIDWOOD (ROCK) s 1 .  :  : p  :  : E 0.9 ......  : ..........................  :,, ....... :* .. *'.*;*-*-:*;

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                         - - NIH PGA*.1                                                 -+- NIH                           PGA*.2                              -3/4- M1 D1                            --B-        M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 2-B BRAIDWOOD (ROCK) s 1  : l p  : ..  : E 0.9 ~ > H: --* ** * **** C T 0.8 R A . .. ..... ..... L 0.7 A 0.6 0 0 0 0 *** 0 H U ... 00 0 0:; 0 O H * .. .. ... .. 0 0. 0 0 00 >< *

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              ~        NIH PGA*.1                                           NIH PGA .2                     11               4 - M1 D1 --B-      M2 D2 NIH  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO"DRAFT RG-1015 FIGURE 3-A FERMI (ROCK) s 1 p .. ... . .. ... E 0.9 .  : .. C T 0.8 .: . *= ** R A 0.7 .... . . . . .;. . .. .. . L 0.6 .. .. . .. ~- . . . . . .* . . ..... ...... ... . . A C C 0.5 . . ...................... . E L 0.4 .. ..... . ... .. . i""' .......... . . . ;, ....: . ; . : . . . ~ .. . . :- . .. .... .... : . . . .. E . . R A 0.3 .... t** . T I 0.2 0 N 0.1 . .. . .... . . g 0 1 10 100 FREQUENCY (HZ)

                     --- NIH PGA*.1                         -+- NIH            PGA*.2                  ---1/4- M1 D1 --B-        M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 3-8 FERMI (ROCK) s 1 .: .: p ... 0.9 .. .... . . .. .. . . .. .. : . .. . . .. .. . . . . . . . .... . .... ..  : ........... : ......... ~ .... ;.. .... E C i : !:- ~. .. .  : T 0.8 . . . . . **:** R . ~  ; A 0.7 ... .. . ***-:* ...... ...... .... .. .....  : ............ : .. . L 0.6  :*** . ..... -: . .:.... . .. . .. . *":". . .... . ..... . i ..................... *********;,,,. : . . . A  : . . . . C . ....... . C 0.5 .. . ... .......

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                        --- NIH PGA*.1                                     -+- NIH                    PGA*.2                       M1 D1                          --B-        M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 4-A COMANCHE PEAK (ROCK) 8 1  :  : . ~  : .

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                        - - NIH PGA*.1                                     -+- NIH                     PGA*.2                        -3/4- M1 D1                       -B-         M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 4-8 COMANCHE PEAK (ROCK) s 1  :  : p E 0.9 ... - . . , ..... *.****** . I... . . :*****'.. . ... C T 0.8 ......... ' ..... R A 0.7 ... .. **********:****** .............. ;.....  :.

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                --- NIH PGA*.1                                -+- NIH             PGA*.2               4--- M1 D1        -&-   M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 5-A SEABROOK (ROCK) s 1 p ~ ~ ...* ... E 0.9 . ****** .* . . .... .. . ...*. . : ... . . . ,. .. .. :

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                        --- NIH PGA*.1                                  -+- N/H             PGA*.2                      ---1/4-         M1 D1                    -B-      M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 5-8 SEABROOK (ROCK) 8 1 p E 0.9 C T 0.8 R A 0.7 L A 0.6 C C 0.6 ********************** * ~ ***** *********r* ******** ***** *! * **!*******:*****"  : * ************** ****** *** * ***** * ****1* *  :

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                       --- NIH PGA*.1                                    -+- NIH             PGAa.2                  ~          M1 D1              --B- M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 6-A RIVER BEND (DEEP SOIL) s 1 p . .. ....... .... .. . . .... ... .. E 0.9 . C T 0.8 . . ..... .. ..... . ... " .. R . . A 0.7 **~*** . . . .........* *****: ............. *: .. . .. ... . . ~ . .... .. . .: .. ..  :. .  :. .. . . : . L . . . i .  : ~ ~  : 0.6 . ..... . .. .... ...... ...... . .....* .. . ' .... :. .......... . A C . . . . . C 0.6 . . . .. .. . .. . . . . ... . .... ... E ' .  : L 0.4 * * * * * ** * * ** * .... ** * * ** * * * * * * ** ~- * ~ * * ** .. . * . .. .. .  :,* .. . * * . }**** ** ** ! * . ** : .. , E  :  ;--=--  : . R  :  : A 0.3 . T . .. ...... . .. .. . .. . ... I 0.2  :  : a N 0.1 . . ... ... . .. g 0 1 10 100 FREQUENCY (HZ)

                    -          NIH PGA*.1                                   -+- NIH                  PGA*.2                       ~         M1 D1                      --B-     M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 6-8 RIVER BEND (DEEP SOIL) s 1 p E 0.9 . . . C T 0.8 R A 0.7 ... . . *=*** . . ... * . .. .. . ....... ... L 0.6 . .  : ..... .. .. . . ... ...... .  : . A C C 0.5 Oo O

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                - - NIH PGA*.1                            -+- NIH          PGAa.2             ---1/4- M1 D1          --B-     M2 D2 N/H  NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 7-A VOGTLE (DEEP SOIL) 1 '  :  :  : s . . p E 0.9 O ******* 00 **

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• NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

RESPONSE SPECTRA COMPARISONS INPUT TO DRAFT RG-1015 FIGURE 7-8 VOGTLE (DEEP SOIL) s 1  :  : . p 0.9 ............................. . E . . C . T 0.8 . . .. ... R . . . A 0.7 *  : ** *** * * *** * * :... * ** * ** * **** ***** ** * ~ * * .. * * * * * *** ** ** * ** ,t. * * **** ....... i. ... ~ .... :.. '" . : .. L A 0.6 ... : ... .  ;. . . .. .. ... ~- ..  : .... *; . . . .~- . . . . . . . ...... .. .. . ....... : ............... i* . . . C C 0.5 E  :  : ~  :  : L 0.4 ...... ;-* . ..  ? .. . *..... : .... :... :- ... . . . . .... * ... . .. * . . . E R .........:.. . A 0.3 * * **** ** *

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                - - NIH PGA*.1                                     -I- N/H PGAm.2                                          -1/4- M1 D1                           -B-        M2 D2 N/H . NEWMARK AND HALL FROM NUREG/CR-0098 ALL SPECTRAL SHAPES ARE 84TH PERCENTILE SHAPES

I. In mak1ng use of both detenninistic and probabilistic evaluations, how should they be combined or weighted, that is should one dominate over the other? The U.S. Nuclear Regulatory Corrmission (NRC) staff feels strongly that deterministic investigations and their use in the development and evaluation of the Safe Shutdown Earthquake Ground Motion should remain in an important aspect of the siting regulations for nuclear power plants for the foreseeable future. The NRC staff also feels that probabilistic seismic hazard assessment methodologies have reached a level of maturity to warrant a specific role in siting regulations.} It is reconnnended that neither detenninistic or probabilistic dominate over the other and that an approach be developed which integrates the two methods.

• II. In making use of the probabilistic and detenninistic evaluations as proposed in Draft Regulatory Guide DG-1015, is the proposed procedures in Appendix C to DG-1015, adequate to determine controlling earthquakes from the probabilistic analysis?

The approach outlined in Appendix C is technically adequate to determine dominate magnitudes and distances. However, the simple detennin~tion of mean magnitude and distance should provide the required information in comparison to the magnitude and distance bins shown in Table C.3 which appears to be very complex. III. In determining the controlling earthquakes, should be median values of the seismic hazard analysis, as described in Appendix C to Draft Regulatory Guide DG-1015, be used to the exclusion of other statistical measures, such as, mean or 85th percentile? (The staff has selected probability of exceedance levels associated with the median hazard analysis estimates as they provide more stable estimates of controlling earthquakes.} It is recorrmended that two probability values be reviewed to derive dominate magnitudes and distances. The two values are:

1. 50th percentile of exceeding current population of reactors based on median hazard curves; and

2. 50th percentile of exceeding current population of reactor based on mean hazard curves considering revised Lawrence Livennore National Laboratory mean hazard results.

IV. The proposed Appendix B to 10 CFR Part 100 has included in Paragraph V(c) a criterion that states: "The annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if , it is less than the median annual probability computed from the current [EFFECTIVE DATE OF THE FINAL RULE] population of nuclear power plants. 11 This is a relative criterion without any specific numerical value of the annual probability of exceedance because of the current status of the probabilistic seismic hazard analysis. However, this requirement assures that the design levels at new sites will be comparable to those

2 at many existing sites, particularly more recently licensed sites. Method dependent annual probabilities or target levels (e.g., lE-4 for Lawrence Livermore National Laboratory or 3E-5 for Electric Power Research Institute) are identified in the proposed regulatory guide. Sensitivity studies addressing the effects of different target probabilities are discussed in the Bernreuter to Murphy letter report. Comments are solicited as to: (a) whether the above criterion, as stated, needs to be included in the regulation? and~ (b) if not, should it be included in the regulation in a different form (e.g., a specific numerical value, a level other than the median annual probability computed for the current plants}? It is reco11111ended to not include the probability numbers as currently specified in the regulation. The regulation should read "The annual probability of exceeding the SSE ground motion is considered acceptably low if it is comparable' to existing reactors which have been shown to be on the order of 10- 3 to 10- 4 per year or lower." V. For the probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power plants? (For the four trial plants used to develop the criteria presented in Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probability of exceedance level. Controlling earthquakes were evaluated for this frequency band, for the average of I and 2.5 Hz spectral responses, and for peak ground acceleration.) At a minimum, 2 response frequencies should be used to determine dominate magnitudes and distances .

DOCKET NUMBER PR t:#\ O OSED RULE..:...:.::...:;:;_*v cl.J-(O 0

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                                                                                        *93 MAY - 3 Al 1 :15
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I( f Florida Power CORPORATION Crystal River Unit 3 Docket No. 50-302 April 23, 1993 Mr. Samuel J. Chilk, Secretary U. S. Nuclear Regulatory Commission Attn: Docketing and Service Branch Washington, D. C. 20555

Subject:

Proposed Rulemaking 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria," (57 Federal Register 47802 - October 20, 1992 and 55601 - November 25, 1992)

Dear Mr. Chil k:

The Nuclear Management and Resources Council {NUMARC} submitted comments on the subject rulemaking in their letter dated March 24, 1993. Florida Power Corporation (FPC} endorses the NUMARC comments. Sincerely,

             ~~:

Nuclear Site Support xc: Mr. William H. Rasin, NUMARC MAY 111993 Acknowledged by card ................._ ~

I V.S. NUCLf:AM :".!1: :-.-~A:QF!Y COMMISSIOt-. ooci-::::: :!~C g_ c:ERJl~E '- ECTION OFF:C[: OF THE "ECRETARY OF THE COMMISSION Document Statistics

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DOCKET NUMBER pn t; 0 S1. f-- \ 00 PROPOSED RULE ft .) - . .. . (_ 5 7 FR J-/ 1 S,-o-i_)  :,,frJJ;/{- [iID *93 ~PR 26 1\11 :16 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 April 22, 1993 Mr. Samuel J. Chilk Secretary of the Commission ATTN: Docketing and Service Branch U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Chi lk :

NUCLEAR REGULATORY COMMISSION (NRC) - REQUEST FOR COMMENTS ON PROPOSED RULEMAKING TO 10 CFR PARTS 50, 52 AND 100, REACTOR SITING CRITERIA TVA has reviewed and is pleased to provide comments on the proposed rulemaking to 10 CFR Parts 50, 52 and 100, Reactor Siting Criteria which was noticed in the October 20, 1992, Federal Register (57 FR 47802). TVA supports the comments on this proposed rule that were submitted by NUMARC on March 24, 1993. TVA appreciates the opportunity to respond to this request for comments. Sincerely,

~ ~ Q~zynski

/J- Manager Nuclear Licensing and Regulatory Affairs cc: See page 2 IAY 111993 Acknowledged b ca y rd......*******...,,,,,,,--- '

1

• ~S. NUCLfAn l1ECULATORY COMMISSlOt-.

COCK:: T,~G & St:AV1CE SECTION OFFICE OF THE SECRETARY Of THE COMMISSION Document StaE5tics

Mr. Samuel J. Chilk Page 2 April 22, 1993 cc: U.S. Nuclear Regulatory Commission Region II 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Dr. Andrew J. Murphy Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555 Mr. William H. Rasin Nuclear Management and Resources Council 1776 Eye Street, NW Suite 300 Washington, DC 20006-2496

ocG~(ET Nrn,rnER

                                                    .- . ... ~-* ~-=r) i:, ; 11 r-PR ro        ~2.. ;.../ 0 0
                                                         *     - .. ~ . t    ._ _      2. j State of Delaware                     f ~ 1 f- fc_ '-/1 f-0:)

DELAWARE GEOLOGICAL SURVEY \..... UNIVERSITY OF DELAWARE d\t. I LJ Newark, Delaware lJ'.) NHC 19716-7501 ROBERT R, JORDAN, STATE GEOLOGIST *93 APR 14 A9 :36 DELAWARE GEOLOGICAL SURVEY BUILDING PHONE: 303*831-2833 FAX: 302*831*357!1 1_1 r, ! 1

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March 10, 1993 Andrew J. Murphy, Chief Structural & Seismic Engineering Branch Division of Engineering Office of Nuclear Regulatory Research Nuclear Regulatory commission Washington, DC 20555

Dear Mr. Murphy:

I wish to offer the following comments on the proposed revision of "Geologic and Seismic siting Criteria for Nuclear Power Plants" sent to us in January. A general comment: It is good to, wherever possible, incorporate state-- of-the-art improvements in flexible regulations. Federal Register Notice 57, No. 203. P. 47810 Urge the adoption of Recommendation 3. There should be reasonable assurance that interdictive measures are possible to limit groundwater contamination resulting from class accidents within the immediate vicinity of the site. (No adequate reason is given for not recommending it.) P. 47820, IV, Required Investigations Urge use of the words "geologic mapping" instead of "geological. ** must be investigated *** " DG-1015, p. 5 "Geological Investigations" should include "geologic mapping." P. 6, lines 13, 14 Acknowledges large source areas for seismic events in the eastern U.S. Therefore, site investigations (geologic mapping) should cover a wider area in the east. Also refer to page 11, lines 10, 11. May need to be larger than radius of 5 miles in the east. MAY 111993--- Acknowfedged by card ...........- ...:............::;

DG-1017, p. 5 What happens to data from seismographs at plants? There should be closer coordination between those responsible for this data and state geological surveys that have seismological programs. For example, closer coordination between the Delaware Geological Survey and the Salem (NJ) Nuclear Power Plant could augment both units' data banks. If recent technological advances cause revision of regulations, should not the advances in data and regulations in the future result in redesign of plants, or at least, methods? DG-4003, p. iii Where does the U.S. Geological Survey fit in? What is its role? P. 3 Urge more than reconnaissance type surveys in the initial stages of site selection. Detailed geologic mapping, such as that provided by state geological surveys should be promoted.

p. 23, line 5 Suggest "must be completed" instead of "should be completed."

In general, we look for increased coordination between State geological surveys and those operating seismic instrumentation at nuclear power plants, plus increased use of geologic mapping provided by geological surveys. Thank you for the opportunity to comment. Sincerely,

                                                ~4();Jm Thomas E. Pickett Associate Director dew

MINISTERE DE L'INDUSTRIE L'Ou~{Ji/cLBUNDESMINISTERIUM FOR ET DU COMMERCE EXTERIEUR UMWELT, NATURSCHUTZ

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Mr SELIN Chairman

• U.S. Nuclear Regulatory Commission WASHINGTON DC 20555 USA OBJECT : Proposed amendments concerning U.S. siting criteria for nuclear power plants.

Referring to the proposed amendments concerning U.S. siting criteria for nuclear power plants, we have some comments on demographic and seismic issues as follows, Demographic criteria US-NRC intends to clearly decouple siting criteria from plant design features. In our meaning, the basis for demographic .. criteria is essentially the possibility to implement efficient emergency measures in case of an accidental situation (evacuation, sheltering, foodstuffs consumption control, ...) ; accordingly, we think that a link must be maintained between demographic criteria and plant design features. Criteria defined for the present generation of nuclear power plants must not be renewed for the next generation of plants without considerations on the type, nominal power and containment characteristics of such plants. We agree that special attention has to be paid to the distances from the plants to cities and/or densely populated areas (and to the evolutions of the demographic characteristics of the sites during the operating life of the plants), as one among the various parameters concerning the preparation of emergency measures. But technically speaking, this problem cannot be dealt with by the means of a single population density limit of 500 persons per square mile up to a distance of 30 miles. Furthermore, the value of 30 miles seems high and not justified. IAY 111993 ~* Acknowledged by card .....................,..H,...,,;:

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It was clearly stated during the OECD-NEA meeting of January 14th, 1993, that the proposed criteria are based upon U.S. experience and are intended to put together elements that have worked well in the past with respect to emergency planning, public acceptance, ... without explicit reference to safety goals or cost/benefit analyses. This point must be very clearly stated in the proposed rules. In addition , such detailed numerical criteria should remain guidelines and not be set as mandatory in formal regulations. II Seismic criteria Concerning the proposed use of both deterministic and probabilistic methods for evaluating the Safe Shutdown Earthquake of a site, we think that improvements in the

,     field of ground motion assessments would mainly proceed from a better knowledge of the physical phenomena and the seismic data (delimitation of seismic sources, expected magnitudes, ... ).

Nevertheless, probabilistic studies can be useful to obtain a better homogeneity of the protection against earthquakes in a country by an appropriate weighting of deterministic safety coefficients. In addition, ii does not seem reasonable to require that the annual probability of exceeding the Safe Shutdown Earthquake ground motion at a site ~e lower than the median annual probability of exceedance computed for the current population of the operating plants. About the new proposed rule concerning the use of the Operating Basis Earthquake for design purposes, it appears that today, in connection with the choice of design rules in existing codes, the design of some safety related components and structures can be strongly influenced by the OBE as well as by the SSE. Considering the safety level improvement required for the next generation of nuclear power plants, a change in the use of the QBE cannot be discussed without considering the whole context of design rules related to QBE and SSE. Walter HOHLEFELDER Director of Nuclear General Director of BMUIRS Safety Installations Copy: Docketing and Service Branch (US NRC)

Florida Power & Light Company, P.O. Box 14000, Juno Beach, Fl 33408-0420 neeKET NUMBER Pl r *TJ-~s::i. ~/ o o March 24, 1993 F;::,OPOSED RULE  :::, , , - L 72 *93 Wt: 30 P ~ :28 (5, (!-fl L/1t;-O-;) Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Docketing and Service Branch Re: Proposed Rule Reactor Site Criteria; including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants 57 FR 47802 (October 20, 1992) Request for Comments

Dear Mr. Chilk:

Florida Power & Light Company (FPL) is the licensed operator of two nuclear power plants in Dade County, Florida and two nuclear power plants in st. Lucie County Florida. FPL has reviewed and endorses the comments submitted by the Nuclear Management and Resources Council (NUMARC) by letter dated March 24, 1993 except as modified and supplemented by the six comments in this letter, all of which pertain to seismic site criteria. comment 1: Grandfather existing sites The effect of the proposed siting criteria will likely be that some sites presently licensed under 10 CFR 100 Appendix A ("Seismic and Geologic Siting Criteria for Nuclear Power Plants") will not meet the proposed new 10 CFR 100 Appendix B ("Criteria for the Seismic and Geologic Siting of Nuclear Power Plants On or After [Effective Date of the Final Rule]") . This would prevent new plants from being built on existing sites. The proposed rule should be revised to include an explicit statement that new plants can be built under Appendix Bon sites presently licensed under Appendix A. The rule should also explicitl y state that the siting requirements of Appendix B will not be applied to license renewal. Comment 2: The application of new OBE-exceedance requirements could lead to changes in "applicable staff posi tions 11 for currently licensed plants Title 10 CFR 100 Appendix A, Section V{a) (2), states, in part, that "If vibratory ground motion exceeding that of the Operating Basis Earthquake occurs, shutdown of the nuclear power plant will be required," but does not explicitly discuss how the NRC will MAY 111993 an FPL Group company Acknowledged by card ......." .....~...'"..._ ..

accomplish this if the plant will not voluntarily shut down. Title 10 CFR 50.36(c) (2) states, in part, that "When a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met," but there are no OBE LCOs in the Technical Specifications for the current generation of nuclear plants. This has not been an issue to date because the two nuclear plants that have experienced OBE exceedances, Summer and Perry, were not operating at the time. Explicit, shutdown criteria are now being proposed in new Appendix s to 10 CFR Part 50, and the associated enforcement authority is being proposed in new paragraph 10 CFR 50.54(ee). Appendix sis applicable only to new plants, therefore, the new authority in 10 CFR 50. 54 (ee) is only applicable to new plants. The application of 10 CFR 50.54(ee) to existing plants would constitute a change in an applicable NRC Staff position and would require a Regulatory Analysis in accordance with the backfit rule (10 CFR 50.109) as part of the proposed rulemaking. Therefore, FPL believes that 10 CFR 50.54(ee) is not applicable to existing plants. Comment 3: Lawrence Livermore National Laboratory (LLNL) NOREG/CR-5250 (January 1989) At a public meeting held by the NRC in Rockville, MD on March 9, 1993, LLNL made a presentation which showed reductions in seismic hazard at nuclear power plants of one to two orders of magnitude below the values published by LLNL in NUREG/CR-5250, "Seismic Hazard Characterization of 69 Nuclear Power Plant Sites East of the Rocky Mountains." These reductions were made possible by a re-elicitation by LLNL of expert opinion in the inputs of "recurrence" and "attenuation". As a consequence of this presentation, it is clear that NUREG/CR-5250 is no longer supported by LLNL. Further, it appears from comments made by LLNL at the meeting that LLNL has no plans to publish a revision to NUREG/CR-5250 to document the reduced seismic hazards it has calculated. This being the case, NUREG/CR-5250 should not be referenced in the proposed rulemaking and its seismic hazard curves should not be used in the 10 CFR 100 Appendix B site selection process. comment 4: Use of the Safety Goal Policy statement to determine the reference probability for the Safe Shutdown Earthquake (SSE) The Safety Goal Policy Statement was approved by the Commission in June 1986 and published in August 1986. On June 15, 1990 the Commission published a staff requirements memorandum (SRM) which offered the following guidance to the NRC Staff regarding implementation of the Safety Goals: 2

0 The staff should establish a formal mechanism, including documentation, for ensuring that future regulatory initiatives are evaluated for conformity with the Safety Goal. 0 A core damage probabilitI4 of less than 1 in 10,000 per year of reactor operations (lx10 ) is a useful benchmark for judging regulations on accident prevention. o Safety Goal objectives should be targets for generic regulatory requirements (see SECY-89-102). The industry's goal of designing future reactors to a core damage probability of less than 1 in 100,000 per year of reactor operation (lxl0- 5 ) was cited by the SRM as evidence of industry's commitment to NRC's severe accident policy. The Commission emphasized that the NRC will not use industry's design objectives as the basis to establish new requirements. The same point was made by the ACRS in correspondence to Chairman Zech 4 dated February 16, 1989. Thus, a core damage frequency of lxl0- /reactor-year is "safe enough, 11 and the NRC Staff should not impose more stringent requirements, even on future reactors, without revising the Safety Goal Policy statement or performing a Regulatory Analysis under the backfit rule. The Safety Goal Policy statement has been considered and used in the non-seismic portions of the proposed rulemaking and there are no reasons to exclude it from consideration and use in the seismic portions of the proposed rulemaking. comments: Probabilistic seismic hazard estimates Probabilistic methods can be used to develop a family of seismic hazard curves which plot peak ground acceleration (PGA) versus probability. In the family are mean values, median values, means plus standard deviations, various percentiles, and so forth. The NRC has chosen to use median values, whereas FPL prefers to use mean values because they are used in the Safety Goal Policy Statement and are preferred in PRA work. To use the seismic hazard curves to develop site specific PGAs, they must be entered at a (generic) "reference probability," which can be derived from the Safety Goal Policy Statement as follows: Assume a fragile plant having neither seismic design nor seismic capability to withstand a seismic event of any size. The probability of core damage then becomes the probability of having the seismic event. Since the Safety Goal Policy Statement sets the threshold for core damage at lxlO -4 , then the acceptable seismic event frequency becomes lxl0- 4 , and the corresponding PGA is obtained from the site specific seismic hazard curve. If the plant is designed and constructed to withstand a seismic event of that PGA, there will be no core 3

damage at the acceptable probability. This is the Safety Goal earthquake level. Applying the foregoing to FPL's Turkey Point site, and using the EPRI seismic hazard curves, design response spectra for the mean, median, and 85th percentile have been plotted (see Attachment 1). These design response spectra have t he shape characteristic of Eastern USA sites where the large "hump" at frequencies less than 10 Hz (as found in current response spectra developed from West Coast data) is missing and the tail of the response spectra extends into what would be the "rigid" range based on West Coast data. This is not a concern because the higher frequencies have been shown to be non-damaging in EPRI NP-5930, "A Criterion for Determining Exceedance of the OBE". Using mean values, the true PGA for FPL's Turkey Point site is about 0.06g, which is 40% of the presently licensed PGA of 0.15g, a value that was establ ished in the 1960s. If the site were to be licensed today using probabilistic methodology, the design response spectrum for the SSE would be a horizontal line at 0. lg over all frequencies, caused by the legal requirement for 0.lg minimum PGA. To summarize, the generic reference probability should be derived from the Safety Goal Policy Statement rather than the methodology in the proposed rule, which ignores the Safety Goa l s and uses median probabilities developed from the un-weighted SSEs of all existing plants regardl ess of design, location, or vintage. The proper selection of the refe r ence probability is of critical importance because it is the one number which all sites must use to determine their site specific PGAs from their site specific seismic hazard curves. A direct link between the reference probability and the Safety Goals is both possible a nd desirable in order to provide legitimacy for the reference probability. comment 6: use of 0.1g as well as 0.3g PGA for future standard plants As discussed above in Comment 5, south-Florida sites have a true SSE in the vicinity of 0.06g PGA (raised to a "regulatory minimum" value of 0.lg PGA for design purposes). Since new standard plants are being designed for 0.3g PGA, the NRC Staff may tend, over time, to think in terms of a 0.3g PGA for all sites. If 0.3g PGA were applied to south-Florida sites, it would result in a PGA five times greater than the true PGA. This would inflate construction costs without any improvement in nuclear safety and would negatively impact the economic analysis needed to justify any commitment by FPL to future nuclear power plant construction. 4

There should be a provision in the new rules for new standard plants to be designed for 0.lg SSE in those areas of the country where the true SSE can be demonstrated to lie below 0.lg. Also, the OBE should not be required to be considered in the design of such plants, nor should shutdown at any level below o. lg be requ ired unless there is significant plant damage. Summary of FPL comments

1. New plants must be permitted on currently licensed sites (grandfathered under proposed 10 CFR 100 Appendix B), and "license renewal" for existing plants must be permitted without consideration of proposed 10 CFR 100 App endix B.

2. The regulatory authority proposed in 10 CFR 50.54(ee) for plant shutdown given OBE exceedance must not be applied to existing plants.

3. NUREG/CR-5250 (January 1989) must not be used or referenced in the proposed rulemaking or supporting regulatory guides.

                                                          -4

4. The same Safety Goal threshold for core damage ( lxl0 /reactor-yr) must be used for future plants as well as currently licensed plants until such time as the Safety Goal Policy Statement is revised to specify otherwise.

5. The reference probability by which the Safe Shutdown Earthquake (SSE) is determined must be derived from and traceable to the Commission's Safety Goal Policy Statement.

6. New standard plants must make provision for a Peak Ground Acceleration (PGA) of 0.lg in their seismic design to accommodate new sites licensed per 10 CFR 100 Appendix B whose true SSE is less than 0.lg. These low seismic sites should not be required to consider the Operating Basis Earthquake (OBE) for design and should not be required to shut down at any level less than 0.lg PGA unless there is significant plant damage.

FPL appreciates the opp ortunity to submit these comments. Vice President Nuclear Engineering & Licensing WHB/JRL/vmg

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Nuclear Dr Ivan Selin Electric Chairman Us Nuclear Regulatory Washington, DC 20555 Nuclear Electric pie Barnett Way Barnwood Gloucester Gl4 7RS Telephone 0452 652222 Telex 43501 Fax 0452 652776 31 March 1993

Dear Dr Selin,

Nuclear Electric's comments on Proposed Changes to 10 CFR parts so, 52 and 100 In response to the 20 October 1992 Notice in the Federal Register, please find attached comments on the NRC's proposed changes to 10 CFR 50, 52 and 100 from Nuclear Electric's perspective in the United Kingdom. The development of common international safety standards are an important goal that should be achieved for the continued safe development of nuclear power throughout the world. We therefore hope that the Commission conniders this goal *together with our comments and other countries perspectives in finalising changes to the existing Regulation. Yours Sincerely,

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Dr B Edmondson Director of Health and Safety Nuclear Electric IAY 111993 Acknowledged by card ..........." ......- ............ Enc: Nuclear Electric's Comments on Proposed Changes to 10 CFR 50, 52 and 100 Registereo office Barnett Way Barnwood Gloucester GL4 7RS Registered in England and Wales Registered No 2264251

Nuclear Electric's Comments on Proposed Changes to 10 CFR 50, 52 and 100 A Reactor Siting Criteria (Non Seismic) The main changes to the proposed regulations are considered below in relation to Nuclear Electric's practice in the Unit~d Kingdom. 1 Exclusion Area Decoupling siting criteria from source terms and dose calculations: The existing US regulation defines the exclusion area based on dose limits at the boundary of this area. To decouple these aspects by setting a very restrictive exclusion area could allow a relaxation in reactor safety to be accepted and place the emphasis on the site itself rather than on the reactor design. In practical terms, if the exclusion area is increased beyond current practice this could permit the potential source terms to be increased whilst still meeting a given dose rate at the boundary. This is not the normal approach in the nuclear industry. In the United Kingdom, for Nuclear Electr ic ' s lic*~nsed sites, an exclusion area for habitation around the site is formed by the site fence. The nearest habitation is then some distance beyond this boundary. For all the existing Nuclear Electric operating sites, the Site Fence is nearer than the distance proposed for the us exclusion zone for future sites of 0.4 miles, 640m. The nearest habitation however is greater than this distance for four of the nine existing sites. Five of the nine existing operating sites do not meet this new criteria for at least one of the reactors on the site. The statutory requirement in the UK relating to Nuclear Electric' s plant is to reduce the normal operation and accident risk from a particular station as low as reasonably practicable (ALARP)

• Acceptability of a particular design at a particular site is then demonstrated by Nuclear Electric by showing that the risk to the individual, taking account of actual demographic data, is acceptably low.

2 Low Population Zone and 7 Feasibility of carrying out Protective Actions It is proposed to decouple the Low Population Zone from site suitability by requiring important site factors such as population distribution, topography, and transportation

routes to be considered to determine whether there are any site characteristics that could pose a significant impedime nt to the development of an emergency plan. These changes would bring the US regulation more in line with UK practice where emergency planning requirements have to be met for each of Nuclear Electric's reactor sites, taking account of the specific reactor design, site and population characteristics: Thesa emerger.cy planning requirements primarily relate to the release associated with the worst design basis accident for the particular reactor design being considered. This determines the radius of the emergency planning zone for which a detailed emergency plan must be in place to demonstrate that the affected sectors within this zone are capable of evacuation within 2 hours. For Nuclear Electric 's Magnox stations this gives a 2-3 kilometer zone and for the AGRs a 1 kilometer zone, with the PWR at a Magnox site currently using the Magnox criteria. The site must be such that the detailed emergency plan should be capable of extension in the event of a more severe accident taking place. 3 Population Density Criteria We agree that current plant designs can and are being shown wo:.::ldwide to have acceptable risks at sites that have significantly higher population densities than those being proposed in the regulation. Hence i f the proposed new criteria are to be used purely to determine whether alternative sites with lower population densities should be considered, this will lead to confusion, particularly outside the nuclear industry and in other countries. If this is the case then we recommend that these values remain i n the Regulatory Guide alone as already suggested as an appropriate alternative. This wou ld be similar to the position applied to Nuclear Electric's sites in the UK where population density requirements have been used as guidelines in determining site suitability: The United Kingdom is densely populated by comparison with the United states. The avera ge population density is 230 per square kilometer, 596 per square mile. This in itself is h i gher than the 500 per square mile population density proposed in the draft regulat ion for any radial distance out to 30 miles. For Nuclear Electric's existing sites the siting criteria are based on limiting criteria for collective dose to the thyroid for a population density specified out to 8 kilometers (5 miles).

The re are currently two sets of siting requirements for the e xi sting Nuclear Electric sit es: one for remote sites, where the early Magnox stations and the first PWR are located, and one for semi-urban sites where the AGRs are located. Population densities all around the site, with limits on the worst 30° sector, are specified. The remote site average population density of 130 per square kilometer, 337 per square mile, is within 500 per square mile. Hence all of Nuclear Electric's existing remote sites meet the proposed us regulation to a distance of 8 kilometers (5 miles). They also meet the proposed regulation to a distance of 15 kilometers (9 miles). The semi-urban site criteria of 1000 persons per square ki l ometer, 2590 per square mile, all around the site and 5000 per square kilometer, 12950 per square mile, in the worst 30° sector are well in excess of the proposed regulation, implying that Nuclear Electric's current AGR sites may not meet the proposed us regulation for future reactors. In fact to a distance of 15 kilometers (9 miles), some sites meet the proposals, others do not. These relaxations in siting requirements for the semi-urban sites were developed following the increased use of nuclear power in the UK, increased knowledge and experience in the design and operation of nuclear plant and improved safety features of the AGRs, coupled with the limited number of remote sites. These requirements were developed following the UK Government ; s review of their safety and siting policy which reaffirmed the principle behind UK safety control that the major contribution to public safety lies in the design, construction and operation standards achieved. For consistency with other proposed changes, the US regulation relating to siting could simply state a requirement to collect populat i on density data for use in other parts of the regulation. Meteorological Factors Decoupling meteorological factors from site suitability: The c hange in regu l ation maintains the requirement to collect and characterise meteorologic~l data representative of the site, but no longer requires this information to be used to determine site suitability. This is discussed above in relation to the proposed change to the definition of the Exclusion zone where radiological dose calculations are no longer required to determine site suitability. summary In finalising changes to existing Regulations, Nuclear Electric recommend that the NRC continue the policy that

has been followed in principle throughout the nuclear industry, of improving reactor design and choosing logical and suitable sites according to the design. Decoupling siting requirements from accident source terms and dose calculations and setting numerical targets for exclusion zones and population density that are not consistent with existing reactor types, locations and levels of safety that have been acceptably demonstrated, appears to remove the incentive to develop safer designs. The proposed changes to decouple siting and design requirements fail to realise the benefits that may be gained by demonstrating the existing margins of safety afforded by the existing siting criteria and reactor designs or by allowing the existing siting criteria to be relaxed. The standards set by the NRC influence the practice in many other countries. It is therefore important that the criteria to be set are based on well understood rationale, taking account of established international practice. B Reactor Siting Criteria (Seismic) There are four significant changes proposed to the regulations each of these is considered below in relation to the current practice applied to Nuclear Electric in the UK.

1) Removal of detailed guidance on site investigations.

In view of the rapid change in the theory and practice of geophysical and seismological site investigation and the variation in the type of techniques suitable between different sites it is appropriate that the regulations should not prescribe the methods to be used. This approach accords with that taken by Nuclear Electric in the United Kingdom and is welcomed by us.

2) Specification of SSE motions in the free field The proposal that SSE motions should be specified in the free field rather than at the foundation locations of the nuclear power plant is a significant step forward in separating the site investigations from the plant design.

It is Nuclear Electric's current practice in the UK to specify Safe Shutdown Earthquake motions in the free field and so we welcome this change.

3) Operating Basis Earthquake Two changes are made here: separating the OBE specification

from the SSE and requiring shut down of the plant if the OBE is exceeded. The regulatory position in the UK for Nuclear Electric's plant is that an OBE is required but that its level is chosen at the discretion of t he licensee subject to the requirement that no safety related plant, system or structure would be impaired by repeated occurrences of ground motions at the OBE level. It is required that the plant be shutdown and inspected whenever the OBE is exce'9ded. We believe that the proposed change in the us rules governing the choice of OBE allows a rational approach to assuring the safety of nuclear plant against lower levels of ground motion than the SSE.

4) The use of probabilistic seismic hazard assessment in determining the SSE The current UK position to which Nuclear Electric's latest plant at a particular site must be assessed is that the SSE

• should be chosen so that it has a probability of being exceeded of no more than one in ten thousand a year. This implies the use of some probabilistic methods and in practice a fully probabilistic hazard assessment is made.

In the past, the outcome of the probabilistic study has been estimates of expected and uniform confidence peak horizontal ground acceleration (pga) in the free field. The chosen expected pga has been coupled with a design response spectrum developed to be representative of the expected range of UK earthquakes and of the ground conditions at the site (soft, medium or hard). More recently Uniform Hazard Response Spectra have been calculated and used for assessment purposes. We welcome the recognition in the proposed rule changes of the value of probabilistic methods of hazard assessment. However there seems little value in performing both deterministic and probabilistic hazard assessment. The information contained in a deterministic assessment is implicit in the probabilistic results, together with an assessment of the likelihood of the deterministic event. It would seem more logical to require an investigation of the sensitivity of the probabil ist ic results to changes in input parameters than to perform a separate analysis of one possibility. The suggested approach to determining the SSE response spectrum using the concept of 'controlling earthquakes' sits awkwardly with the preceding probabilistic hazard analysis. It would be more logical to follow through the probabilistic analysis to develop a uniform hazard response spectrum with the desired exceedance probability. This would enable the conservatism in the chosen SSE motion to be quantitatively assessed.

summary The removal from the regulation of prescriptive guidance on geological and seismologica l site investigations, the definition of SSE motions in the free field and the transformation of the OBE into a shutdown limit decoupled from the design earthquake are all welcomed as positive steps towards a more rational basis for the safety of nuclear power stations against earthquakes. The adoption of probabilistic methods for hazard assessment is also welcomed since it allows the best available estimates of seismic hazard to be used in the design of nuclear power plant. It is not clear why the probabilistic approach is not to be followed to its logical conclusion in the determination of uniform hazard response spectra. Performing a separate, add itional deterministic hazard assessment seems unnecessary.

6 South Carolina Electric & Gas Company John L. Skolds P.O. Box 88 Vice President

                                                  ~     KET NUMBER SCE&G Jenkinsville, SC (803) 345-4040 PROP~S~ D RULE PR                  Nuclear Operations SO, 52 >- lOD.

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March 24, 1993 Ref er to : RC-93-007 4 *g3 MAR 30 A1 C:18 Mr. Samuel J. Chilk Secretary, U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

Dear Mr. Chilk:

Subject:

VIRG IL C. SUMMER NUCLEAR STATION DOCKET NO. 50/395 OPERATING LICENSE NO. NPF-12 COMMENTS ON PROPOSED RULE CHANGE TO 10 CFR 100: REACTOR SITE - CRITERIA (NUM 920014) South Carolina Electric & Gas Company (SCE&G} has participated with the NUMARC Seismic Ad Hoc Advisory Committee (AHAC} in the review of the proposed 10 CFR 100 Rulemaking package (contained in 57 FR 47802 and 57 FR 55601) and endorses the positions and comments submitted by NUMARC on behalf of the industry. In addition, SCE&G has the following supplementary comments on specific portions of the package. A. DRAFT REGULATORY GUIDE DG-1016

1. The requirement for a free-field instrument may be useful from an academic standpoint, but it introduces numerous uncertainties i n recorded ground motions which mu st be reconciled with the motions recorded insi de plant structures. Recorded motions should be obtained at locations coinciding with the plant design input not at a free-field site which has no design basis for comparison. The requirement for a free-field instrument will impose an undue burden on licensees due to the time required for and the cost involved in the reconciliation process.

2. The number of seismic instruments specified in the Regulatory Guide is excessive, especially in light of industry initiatives and the NRC's concern for limiting radiation exposure and improving operation and maintenance of the instruments. Also, only a limited number of instruments should be used in order to simplify the comparison of recorded motions to the structural design models.

The number and location of instruments recommended by NUMARC appear to be reasonable and address most of these cdncerns. However, SCE&G recoD1r1ends that a free-field instrument not be requ i red as discussed in Comment 1. MAY 11 1993 ., Acknowledged by card .............~.~..................

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. Mr. Samuel J. Chilk NUM 920014 Page 2 of 2 B. DRAFT REGULATORY GUIDE DG-1017

1. The determination of Operating Basis Earthquake (QBE) exceedance should be based on the EPRI NP-5930 criterion as measured by an instrument located on the containment foundation, the location of the plant design input where a direct response spectrum comparison can be made. An instrument located in the free-field provides data that is not directly comparable to the plant design which can complicate/prolong the plant shutdown/restart decision and in general amplifies the recorded motions significantly over what the plant actually experiences.

If you have any questions, please call at your convenience. Very truly yours,

                                        ~&~

John L. Skolds ARR:lcd c: 0. W. Dixon R.R. Mahan R. J. White G. F. Wunder NRC Resident Inspector R. B. Whorton NUMARC NSRC File (811.02 - 50.071) RTS (NUM 920014) NUCLEAR EXCELLENCE - A SUritlER TRADITION!

DOCKET NUMBER Pl 5o 'ik f--10 ~ PROPOSED RULE cs 7 FR. '-/7 Jro:,.. j ~ .. V GULF STATES UTILITIES CO.JWPA.N~ - RIVER BEND STATION POST OFFICE BOX 220 ST FRANCISVILLE, LOUISIANA 70775 *93 HAR 30 A10 :17 AREA CODE 504 635-6094 346-8651

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March 24, 1993 RBG- 38 l 2 65 File No. G9.23.1 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Proposed Rulemaking 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria" (57 Federal Register 47802 - October 20, 1992 and 55601 - November 25 , 1992) Gentlemen: Gulf States Utilities Company (GSU) has reviewed the NRC ' s proposed rulemaking concerning Reactor Siting Criteria and endorses those comments provided by the Nuclear Management and Resources Council (NUMARC). Sincerely, 2-r. ~ J.E. Booker Manager - Safety Assessment and Quality Verification River Bend Nuclear Group MAY 1 1 19QJ:: Acknowledged by card ..................................

U.S. NUCU:An Rd),.,i.,-ffORY COMMISSfQr,.. DOCKETING & SERVICE SECTION OFflCE OF THE SECRETARY OF THE COMMISSION Dorument Statistics Pcstrnaik Date /;;,._~ /C)J Spe~iai Distri tion ll-:r, OS:7 P/J((;

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202-9 5 5 - 66 00 March 23, 1993 Office of the Secretary U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Chief, Docketing and Service Branch Re: Proposed Rule on Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition for Rulemaking From Free Environment, Inc. et al. (57 Fed. Reg. 47,802 (Oct. 20, 1992))

Dear Sir:

On October 20, 1992, the Nuclear Regulatory Commission (NRC) published a Notice of Proposed Rulemaking and Proposed Denial of Petition for Rulemaking on "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants." Although foreign utilities are not legally bound - ;by the proposed rule, their national nuclear standards are consistent with the nuclear safety standards of the International Atomic Energy Agency (IAEA), which were strongly influenced by the NRC's siting standards. If the proposed revisions to the siting regulations in 10 C.F.R. Part 100 are adopted, the process for selecting new nuclear power plant sites would fundamentally change, thereby forcing reconsideration of IAEA and national nuclear safety siting standards and raising questions about the adequacy of present and future nuclear power plant sites to ensure adequate protection of the public health and safety in foreign countries. Therefore, foreign utilities have an interest in staying abreast of developments in

• a ~ a and participating 1 1~93 Acknowledged by card ................- .........."...

NEWMA N & liOLTZINGER, P. C, Office of the Secretary March 23, 1993 Page 2 in this rulemaking proceeding due to the potential impact that it could have on reactor siting worldwide. Although agency regulations do not require a formal notice of appearance, we are enclosing for filing the original and two copies of the Notice of Appearance of William O. Doub, L. - Manning Muntzing, and myself, as counsel for the following foreign utilities in this rulemaking proceeding: Atomic Energy of Canada, Ltd. Electricite de France The Federation of Electric Power Companies The Hokkaido Electric Power Co. The Tohoku Electric Power Co. The Tokyo Electric Power Co. The Chubu Electric Power Co. The Hokuriku Electric Power Co. The Kansai Electric Power Co. The Chugoku Electric Power Co. The Shikoku Electric Power Co. The Kyushu Electric Power Co. Taiwan Power Company. Additionally, I have enclosed an extra copy of this letter and the three Notices of Appearance. Please acknowledge receipt of this document by file stamping "Received " on this copy and return the file-stamped copy in the self-addressed, stamped envelope which is provided. erely, J',£7V-- anet E.B. Ecker JEBE/pg Enclosures

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

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Proposed Rule on Reactor Site Criteria; ) Including Seismic and Earthquake ) Engineering Criteria for Nuclear Power ) Proposed Denial of Plants and Petition ) for Rulemaking From Free Environment, ) Inc., et al, (57 Fed. Reg. 47,802 ) (Oct. 10, 1992)) ) ____________________ )

                                         )

NOTICE OF APPEARANCE OF COUNSEL Notice is hereby given that Janet E.B. Ecker enters an appearance of Counsel in the above-captioned rulemaking proceeding for: Atomic Energy of Canada, Ltd. Electricite de France The Federation of Electric Power Companies The Hokkaido Electric Power Co. The Tohoku Electric Power Co. The Tokyo Electric Power Co. The Chubu Electr ic Power Co. The Hokuriku Electric Power Co. The Kansai Electric Power Co. The Chugoku Electric Power Co. The Shikoku Electric Power Co. The Kyushu Electric Power Co. Taiwan Power Company Name: Janet E.B. Ecker Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036 Telephone: (202) 955-6600 Admissions: Court of Appeals for the District of Columbia Name of Party: Atomic Energy of Canada, Ltd. 2251 Speakman Drive Mississuaga, Ontario LSK 1B2 Canada

Electricite de France 22/30, avenue de Wagram 75382 Paris Cedex 08 France The Federation of Electric Power Companies Keidanren Kaikan 1-9-4, Ote-machi, Chiyoda-Ku Toyko, Japan Taiwan Power Company 20th Fl oor 242 Roosevelt Road Section 3 Taipei, Taiwan Repr of China Ja t E.B. Ecker

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                                          )

NOTICE OF APPEARANCE OF COUNSEL Notice is hereby given that William o. Doub enters an appearance of Counsel in the above-captioned rulemaking proceeding for: Atomic Energy of Canada, Ltd. Electricite de France The Federation of Electric Power Companies The Hokkaido Electric Power Co. The Tohoku Electric Power Co. The Tokyo Electr ic Power Co. The Chubu Electric Power Co. The Hokuriku Electric Power Co. The Kansai Electric Power Co. The Chugoku Electric Power Co. The Shikoku Electric Power Co. The Kyushu Electric Power Co. Taiwan Power Company Name: William O. Doub Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036 Telephone: (202) 955-6600 Admissions: U.S. Court of Appeals for the District of Columbia and Fourth Circuits; Court of Appeals for the District of Columbia; Court of Appeals of Maryland Name of Party: Atomic Energy of Canada, Ltd. 2251 Speakman Drive Mississuaga, Ontario L5K 1B2 Canada

Electricite de France 22/30, avenue de Wagram 75382 Paris Cedex 08 France The Federat ion of Electric Power Companies Keidanren Kaikan 1-9-4, Ote-machi, Chiyoda-Ku Toyko, Japan Taiwan Power Company 20th Floor 242 Roosevelt Road Section 3 Taipei, Taiwan Republic o f China ~~~ aino.oouh Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C.d:0036 Dated: c..3 /

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Proposed Rule on Reactor Site Criteria; ) Including Seismic and Earthquake ) Engineering Criteria for Nuclear Power ) Proposed Denial of Plants and Petition ) for Rulemaking From Free Environment, ) Inc., et al, (57 Fed. Reg. 47,8 02 ) (Oct. 10, 1992)) ) _____________________ )

                                          )

NOTICE OF APPEARANCE OF COUNSEL Notice is hereby given that L. Manning Muntzing enters an appearance of Counsel in the above-captioned rulemaking proceeding for: Atomic Energy of Canada, Ltd. Electricite de France The Federation of Electric Power Companies The Hokkaido Electric Power Co. The Tohoku Electr ic Power Co. The Tokyo Electric Power Co. The Chubu Electric Power Co. The Hokuriku Electric Power Co. The Kansai Electric Power Co. The Chugoku Electric Power Co. The Shikoku Electric Power Co. The Kyushu Electric Power Co. Taiwan Power Company Name: L. Manning Muntzing Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, D.C. 20036 Telephone: (202) 955-6 600 Admissions: U.S. Supreme Court; U.S. Court of Appeals for the District of Columbia Circuit; Court of Appeals for the District of Columbia Name of Party: Atomic Energy of Canada, Ltd. 2251 Speakman Drive Mississuaga, Ontario L5K 1B2 Canada

Electricite de France 22/30, avenue de Wagram 75382 Paris Cedex 08 France The Federation of Electric Power Companies Keidanren Kaikan 1-9-4, Ote-machi, Chiyoda-Ku Toyko, Japan Taiwan Power Company 20th Floor 242 Roosevelt Road Section 3 Taipei, Taiwan Republic of China L. Manning ,5 Newman & Holtzinger, P.C. 1615 L Street, N.W. Suite 1000 Washington, o.c. 20036 Dated:~~ ~3) J<:J1?,

-===- ENTERGY DOCKET NUMBER r,r-~1::-- ~~o RULE Pl 5 eI s ..i, d-l otJ (!;1 F~ L/7J--01..) COC'\i:1i:D us, *:; Entergy Operations, Inc. P.O. Box 31995 Jackson, MS 39286-1995 Tel 6019849740

                                                         *93 MA , 29 p _-:; :4g           John R. McGaha March 23, 1993 Mr. Samuel J. Chilk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 ATTENTION:             Docketing and Service Branch

Subject:

Entergy Operations Comment on Proposed Rulemaking to 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria"

Reference:

57 Federal Register47802- October 20, 1992, and 55601 - November 25, 1992 CNRO-93/00016

Dear Mr. Chilk:

The referenced Federal Registers requested comments on the subject proposed rulemaking. Entergy Operations, Inc., the licensee for Arkansas Nuclear One, Grand Gulf Nuclear Station, and Waterford 3 Steam Electric Station wishes to offer the following: Entergy Operations participated in the development of comments being submitted by NUMARC and endorses those comments on this proposed rule. In addition to those comments, we wish to emphasize some points on the proposed rule where it deals with non-seismic siting criteria. We recommend that radiological dose consequence evaluation factors contained in the current 10 CFR 100 be retained as the key determinants for site suitability. There does not appear to be adequate technical justification for arbitrarily mandating the proposed minimum exclusion area size and population density. We believe the proposed change has the potential for significant negative impact to both currently licensed and future plants without appreciable improvement to public health and safety. A k led lffl' 11.1993 c now ged by card .............................,:::;-

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Entergy Operations Comments on Proposed Rulemaking to 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria" CNRO-93/00016 Page 2 of 2 March 23, 1993 A petition to incorporate minimum exclusion area and low population zone distances and population density limits into the regulations has been denied once by the NRC (53 FR 50232) on the basis that it unnecessarily restricted NRC's regulatory siting policies without resulting in a substantial increase in the overall protection of the public health and safety. We concur with the basis of that denial. The background for this proposed rule states the NRC is proceeding because of "possible renewed interest." While there is much interest in reactor siting, adoption of the proposed criteria may adversely affect public perception as to the acceptable safety of existing plant sites. Additionally, it could lead to disqualifying a significant number of existing sites as well as new sites from hosting additional or new nuclear reactors apart from any demonstrable safety consideration. Detailed comments and responses to the NRC questions are included in the attachment. In summary, we believe the criteria contained in the current Part 100, has been and remains appropriate for providing the appropriate protection of public health and safety. Therefore, we respectfully request the NRC to reconsider proposed non-seismic rule. Sincerely, JRM/baa attachment cc: Mr. R P. Barkhurst Mr. J. J. Fisicaro Mr. J. W. Yelverton Mr.RF. Burski Mr. W. K. Hughey Corporate File [ 8 ] Mr. W. T. Cottle Mr. L. W. Laughlin DCC(ANO) Mr. J. G. Dewease Mr. M. J. Meisner Records Center (W-3) Central File (GGNS) CNR0-92/00016 March 22, 1993 Page 1 of 5 Entergy Operations Comment on Proposed Rulemaking (Non-Seismic) to 10 CFR Parts SO, 52, & 100 General We recommend that radiological dose consequence evaluation factors contained in the current 10 CFR 100 be retained as the key determinants for site suitability. There does not appear to be adequate technical justification for arbitrarily mandating the proposed minimum exclusion area size and population density. The proposed codification of population density and minimum exclusion areas size for siting future nuclear power plants does not appear to have any inherent appreciable improvement to public health and safety. However, it could have a negative impact by possibly resulting in the inappropriate disqualification of a more favorable site in preference to a site that is less desirable overall but meeting population density and minimum exclusion areas size requirements. Further, this could lead to negative public perception regarding the safety of existing plants as well as impacting construction of future units at an existing site. Finally, we believe the proposed codification of population density and minimum exclusion areas could send an inappropriate message regarding the safety risk associated with advanced light water reactor designs in general. The proposed rule codifies very conservative numeric criteria for population density and minimum exclusion areas size as key indicators of site suitability regarding offsite radiation dose risk with negligible improvement in protection of public health and safety despite extensive siting experience to date that demonstrated current requirements provide very conservative criteria for site suitability. Based on the industry study Evaluation of Population Distribution Relative to Meeting the Quantitative Health Objectives of the NRC Safety Goal Policy for Offsite Risk Associated with Nuclear Power Plants, we believe the proposed rule would be unnecessarily restrictive and is contrary to the intent of the NRC's Safety Goal Policy. NRC Questions Question 1: Should the Commission grandfather existing reactor sites having an exclusion area distance less that 0.4 miles for the possible placement of additional units, if those sites are found suitable from safety consideration?

Response

The fact that existing sites have been evaluated for suitability from safety consideration apart from the proposed exclusion area and found acceptable is indicative of the problem with this proposed rule. The proposed basis for determining site suitability restricts NRC flexibility unnecessarily

Attachment 1 CNR0-92/00016 March 22, 1993 Page2 of5 with no appreciable increase in public health or safety. The key factors for determining site suitability for additional units at an existing site or evaluating new sites are the radiological dose

• consequence evaluation factors in the current 10 CFR 100. Dual siting safety standards are inappropriate and should be discouraged.

Question 2: Should the exclusion area distance be smaller that 0.4 miles (640 meters) for plants having reactor power levels significantly less than 3800 Megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example, 0.25 miles or 400 meters)?

Response

The appropriate method for determining the exclusion area distance should be determined based on radiological dose consequence evaluation contained in the current 10 CFR 100. Exclusion area distances less than the 0.4 miles proposed have been found by the NRC to be adequate for the protection of public health and safety for approximately one third of the currently licensed operating sites. The flexibility to choose a site based on all factors relating to public health and safety as is currently the case should be maintained. Question 3: The Commission proposed to codify the population density guidelines in Regulatory Guide 4. 7 which states that the population density should not exceed 500 people per square mile out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter. Comments are specifically requested on question 3A, 3B, and 3C given below. Question 3A: Should numerical values of population density appear in the regulation or should the regulation provide merely general guidance, with numerical values provided in a regulatory guide?

Response

Since population density limits are not key determinants of offsite radiological dose risk, they provide essentially no beneficial contribution to the protection of public health and safety regarding offsite radiological dose risk beyond the immediate area adjacent to the power plant. However, in general, regulations should provide regulatory requirements with specific suggested guidance in regulatory guides. In this manner, licensees maintain the flexibility to use alternative NRC approved methods to meet the requirements of a rule. CNR0-92/00016 March 22, 1993 Page 3 of5 Question 3B: j Assuming numerical values are to be codified, are the values of 500 persons per square mile at the time of site approval and 1000 persons per square mile 40 years thereafter appropriate? If not, what other numerical values should be codified and what is the basis for their values?

Response

We do not believe there is adequate technical basis for mandating any minimal numerical criteria apart from its clear link to o:ffsite radiological dose risk and a commensurate benefit to public health and safety. Further, we are not aware of any accurate method to predict population density, growth, and distribution for a period of forty years into the future. Question JC: Should population density criteria be specified out to a distance other than 30 miles (50 km), for example, 20 miles (32 km)? If a different distance is recommended, what is its basis?

Response

See our response to Question 3B above. Question 4: Should the Commission approve sites that exceed the proposed population values of 10 CFR 100 .21 and if so, under what conditions?

Response

The key determinant for site suitability is by using the current radiological dose consequence evaluation factors. See our response to Question 3B above. Question 5: Should holders of early site permits, construction permits, and operation license permits be required to periodically report changes in potential offsite hazards (for example, every five years within 5 miles)? If so, what regulatory purpose would such reporting requirements serve?

Response

Such a reporting requirement for operation licensees (OL) would be redundant to reporting requirements of 10 CFR 50.71(e), and as such would not be necessary or appropriate. For early site permits (ESP) or construction permits (CP), there is no regulatory purpose for periodically CNRO-92100016 March 22, 1993 Page 4 of 5 reporting changes in potential offsite hazards since there are no public health and safety effects due to construction or siting. The proper time to consider such changes would be when the NRC grants an OL or a combined operating license (COL). Since there are effective regulatory requirements in place, we believe this redundant reporting requirement would present an unnecessary burden on both NRC and licensee resources. Further, this would not be consistent with the purpose of the Paperwork Reduction Act of 1980. Question 6: What continuing regulatory significance should the safety requirements in 10 CFR 100 have after granting the initial operation license or combined operation license under 10 CFR 52?

Response

10 CFR 100 Reactor Site Criteria should remain a regulation for evaluation of the suitability of proposed sites. As such, it should not have any continuing regulatory significance beyond the issuance of site permits or siting portion of combined operation licenses. Construction and operational safety requirements are adequately addressed on other parts of the Commission's regulations (e.g. 10 CFR 50). Question 7: Are there certain site meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are the conditions that can not be adequately compensated for by design features?

Response

We are not aware of any meteorological conditions that can not be adequately compensated for by design features. As such, Regulatory Guide 1.145 should be revised to delete the requirement to collect one year of meteorological data since this data is not needed for determination of site suitability. Meteorological investigations to characterize remaining severe weather phenomena for potential applicability and conservative Chi/Q values can be achieved using information that is readily available rather than by collecting site specific meteorological data for an entire year. Question 8: In the description of the disposition of the recommendations for the Siting Policy Task forces report (NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to reconsider any recommendations not adopted and, if so, what are the bases for reconsideration? CNR0-92/00016 March 22, 1993 Page 5 of 5

Response

We believe the elements of this report have been adequately addressed. Additional Comments for Oarification of Appendix Q Proposed revision to 10 CFR 52, Appendix Q provides an applicant for renewal of an early site permit "is subject to a full early site permit review." This proposed revision is inconsistent with the current 10 CFR 52, Section 52.31 as well as Subpart A of part 52. Additionally, this provision does not appear to be applicable for an early site permit renewal because an early site permit holder seeking renewal would be per Section 52.29 and 52.31 rather than seeking a separate review by NRC staff on site suitability under Appendix Q. We recommend the proposed change be deleted based upon the above since it would serve only to undermine the stability and predictability of the siting process.

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                                                                                    *93 MA\: 26 P5 :49 PWM93106                                                                              l \ ._,

March 23, 1993 Mr. Samuel J. Chilk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Proposed Rulemaking 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria" (57 Federal Register 47802--0ctober 20, 1992 and 55601--November 25, 1992)

Dear Mr. Chilk:

General Electric Nuclear Energy (GENE) has reviewed both t he proposed rule, "10CFR Parts 50, 52, and 100, Reactor Siting Criteria" (Federal Register 47802--0ctober 20, 1992 and 55601--November 25, 1992) and the NUMARC comments related thereto. Based upon these evaluations GENE strongly endorses the NUMARC comments and urges the Commission to reflect them in the final version of the revised rule. We particularly endorse the NUMARC concern over the potential impact the proposed rule might have in the siting of future reactors domestically as well as their siting in foreign countries. Si ncerely, rt P. W. Mar r 1 tt, Manager Safe t y .ind icensing M/C 444 408-925-6948 cc: W.H. Rasin (NUMARC) IAY 1 1 1993--

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( 5'1 FR. ~18-6i} WINSTON & STRA FREDERICK H. WINSTON (1853-1886) 1400 L STREET, N.W. CHICAGO OFFICE SILAS H. STRAWN (1891-1946) WAS H INGTON , D.C. 20005-350 2 ".J WEST WACKER DRIVE ICAGO, ILLINOIS 60601 (312) 558-5600 (20 2 ) 371-5700 FACSIMILE (202) 371-5950 75 WATER STREET WRITER'S DIRECT DIAL NUMBER N YORK, NY 10038-4981 (212) 269-2500 March 24, 1993 Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission *93 t'\~R 24 P3 :52 Washington, D.C. 20555

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Attention: Docketing and Service Brane fr' , , *

• RE: Response to Proposed Rulemaking -- Reactor site Criteria; Including Seismic and Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition for Rul emaking from Free Environment, Inc. et al.

57 Fed. Reg. 47,802 (October 20 1 1992) The Nuclear Regulatory Commission ( "NRC" or "Commission") has published a proposed rule that would, inter alia, amend 10 C.F.R. Part 100 to include a new Subpart Band Appendix B, which define a new set of requirements for siting new power reactors. 57 Fed. Reg. 47,802 (October 20, 1992), comment period extended, 58 Fed. Reg. 271 (January 5, 1993). The proposed rule raises an issue of considerable importance to existing Part 50 power reactor licensees -- the continued adequacy of a reactor site previously reviewed and approved by the NRC based on the requirements in effect at the time of licensing. These comments are limited in scope and focus on the possible adverse effect that proposed Subpart B would have on existing reactor sites. To counter that effect, adoption of one of fol l owing two alternatives is required: (1) permit existing reactor sites the option of following either Subpart A or proposed Subpart B for new reactor licensing applications; or (2) revise Subpart B to specifically exempt existing reactor sites from the more stringent requirements associated with t,he exclusion area distance. These comments are submitted on behalf of Northeast Utilities and Washington Public Power supply system.

1. An Applicant to Construct a New Plant on a Reactor Site Previously Reviewed and Approved by the NRC Should Have the Option of Applying Either Subpart A or B to 10 C.F.R. Part 100 The notice states that the proposed rule is prospective, applying only to future Part 50 and Part 52 applicants, not to MAY 111993 ~-

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WINSTON & STRAWN U.S. Nuclear Regulatory Commission March 24, 1993 Page 2 existing reactor sites. 57 Fed. Reg. at 47,803. The language of the proposed rule attempts to implement that intent. For instance, the new siting requirements in proposed Subpart B to Part 100 are to apply to a Part 50 or Part 52 application filed after the effe ctive date of the rule change. See 10 C.F.R. Part 100, prop osed Subpart B (title) and proposed Appendix B (title and "General Information"). However, proposed Subpart B also would apply to an application for the construction of a new plant at a reactor site already reviewed and approved by the Staff. If the new plant were to be located on a reactor site at which another unit (or units) is currently operating or under construction, two separate siting requirements would be applied to the single reactor site -- i.e., Subpart A for the unit(s) currently on the site and proposed Subpart B for the new plant(s) proposed for the site. Because, under such a scenario, the reactor site would have to be re-reviewed and re-approved, the proposed rule is not truly prospective. If the Staff applies the new requirements in proposed Subpart B to a previously reviewed and approved reactor site, and then rejects a new plant to be located there on the basis of inadequate siting characteristics, additional time, effort, and cost would need to be expended to identify and study another reactor site -- and there is not an abundance of sites. The new rule would ignore the fact that a previously approved site had already been dedicated to power production and thereby is likely to have fewer environmental, construction, and acceptance issues associated with it. Moreover, it is unclear what effect the rejection would have on the previously reviewed and approved reactor site. In sum, by precluding the use of a reactor site based on new siting regulations when that site was previously reviewed and approved under an earlier version of the regulation, the proposed rule works an unfair, inequitable, and possibly incorrect result that should be avoided.Y To preclude such an anomalous result, and to ensure that the rule is truly prospective, we recommend that the Staff permit an applicant to construct a new plant on a reactor site previously reviewed and approved by the NRC the option of applying either Subpart A or proposed Subpart B of revised 10 C.F.R. Part 100. Y Cf. Northern Indiana Pub. Serv. Co. v. Porter County Chapter of the Izaak Walton League of America. Inc., 423 u.s. 12, 15-18 (1975) (Douglas, J., noting the danger of retroactively applying an amendment to 10 C.F.R. Part 100 when the pre-existing regulation had already been specifically applied).

WINSTON & STRAWN U.S. Nuclear Regulatory Commission March 24, 1993 Page 3 In addition to these potential considerations, there is a legal implication as well -- backfitting. We conclude that application of the proposed Subpart B siting requirements to approved reactor sites (as opposed to a new plant on the site) may in certain circumstances constitute backfitting within the meaning of 10 C.F.R. § 50.109(a) (1). This conclusion stems from the fact that the Staff can, if presented with an application, complete an early site review under Appendix Q to 10 C.F.R. Part 50 or issue an early site permit under Subpart A to 10 C.F.R. Part 52. If such an approval had been granted, and the reactor site was later rejected for failure to meet the new siting requirements contained in proposed Subpart B, the Staff would have modified its prior approval. Therefore, if our recommendation is not adopted, the Staff should revise its backfitting analysis (see 57 Fed. Reg. at 47,813) to address potential adverse effects on such previously reviewed and approved reactor sites. Finally, the Commission should expressly state in the Part 100 rulemaking that the new siting requirements in proposed Subpart Bare not applicable to the issuance of a renewed license pursuant to 10 C.F.R. Part 54. The licensing of existing sites under the current regulations in Part 100 is specifically encompassed within the "current licensing basis," as that term is defined in 10 C.F.R. § 54.3. Therefore, the subject rulemaking should be amended as suggested in order to preclude any confusion over this fact.

2. Existing Reactor Sites with an Exclusion Area Distance of Less than 0.4 Miles Should Be Able to Place New units at the site Recognizing the potential legal implications attendant to the application of new siting requirements to existing reactor sites, the Staff posed the following question in the proposed rule:

   "Should the Commission grandfather existing reactor sites having an exclusion area distance of less than 0.4 miles (640 meters) for the possible placement of additional units, if those sites are found suitable from safety consideration [sic)." 57 Fed. Reg. at 47,811.

For the reasons developed below, we answer this question in the affirmative. However, we do not agree with the qualifying clause in the question implying the need to re-assess existing site suitability from a safety perspective. If an existing reactor site has an exclusion area distance of less than 0.4 miles, the site should be grandfathered under the proposed rule without any additional review. Should the Staff conclude otherwise, as implied by the last clause in the question, a detailed basis should be provided for following such an approach and the "safety consideration[s)" should be identified that will be used to re-assess site "suitability."

WINSTON & STRAWN U.S. Nuclear Regulatory Commission March 24, 1993 Page 4 Propos ed Section 100.21(a) (1) prescribes a minimum exclusion area distance of 0.4 miles. As the Staff is aware, of the more than 75 power reactor sites approved by the NRC, 25 sites have exclusion area distances of less than o. 4 miles but still satisfy the current Part 100 dose limitations (i.e., Subpart A).Y However, because the proposed requirement at Section 100.2l(a) (1) affords no leeway on the exclusion area distance, these 25 approved reactor sites would not meet proposed Subpart B if they were identified in an application for a new plant. Such a result would serve little purpose other than to foreclose an otherwise acceptable site from future use. The proposed requirement of 0.4 miles for the exclusion area distance stems from the existing Staff guidance identified in Regulatory Guide 4.7. However, proposed Section l00.2l(a)(l) is silent on the option also provided in Regulatory Guide 4.7 of using compensatory pla nt de sign features where the e xclusion a rea distance is less than 0.4 miles.~ While we recognize the Staff's desire in this rulemaking to decouple plant design from reactor siting (relocating the former to Part 50 and retaining the latter in Part 100), to adopt an already arbitrary standard of 0.4 miles without also adopting its counterpart alternative is not in keeping with past Staff practice or Commission policy.Y Moreover, in this proposed rulemaking, the Staff specifically declined to adopt a recommendation that sites should have "no unfavorable characteristics," concluding instead that "applicants may provide specific plant design features to compensate for site inadequacies." 57 Fed. Reg. at 47,810. Y See SECY-92-215, "Revision of 10 CFR Part 100, Revisions to 10 CFR Part 50, New Appendix B to 10 CFR Part 100 and New Appendix S to 10 CFR Part 50," dated June 12, 1992, at 5.

   ~    Historically, the Staff guidance called for reactor sites with an exclusion area of 0.4 miles, but permitted lesser distances where "special conditions on the station design (e.g., added engineered safety features)" were used to meet the offsite doses limitations in Part 100.      See Regulatory Guide 4. 7, "General Site Suitability Criteria for Nuclear Power Stations," Rev. 1 (Nov. 1975), Regulatory Position C.3.

Y See Public Interest Research Group, DPRM-88-5, 28 N.R.c. 829 (1988). There, a petition to change Part 100, including a request to adopt a specific numerical limit of 0.4 miles for the exclusion area distance, was denied on the basis of the guidance in Regulatory Guide 4.7, and the generally recognized need for "regulatory flexibility" when applying the Part 100 siting requirements. Id. at 832-33.

WINSTON & STRAWN U.S. Nuclear Regulatory Commission March 24, 1993 Page 5 Therefore, the proposed requirements in Section 100.2l(a) should be modified as follows (additional language is bolded, deleted language is s:tr-ueJE :through): (a) Each reactor facility must have an exclusion area, as defined in§ 100.J(a) of this part. (1) For sites with a single reactor facility, the distance to the exclusion area boundary at any point (as measured from the reactor center point) shall should be at least 0.4 miles but if the distance is less than 0.4 miles it may be necessary to place special conditions on the station design (e.g., added engineered safety features). (2) For sites with multiple reactor facilities, consideration must be given to the following: If the reactors are independent to the extent that an accident in one reactor would not initiate an accident in another, the size of each exclusion area must be determined with respect to each reactor individually and if the distance is less than o.4 miles it may be necessary to place special conditions on the station design (e.g., added engineered safety features). The exclusion area for the site must then be taken as the plan overlay of the sum of the exclusion areas for each reactor * * * . This proposed language is the same as that contained in Regulatory Guide 4.7, Rev. 1, Position C.3, at p. 4.7-9. In addition, this proposed change is similar to the alternate provision contained in proposed Section 100.21(b) (1)~ which addresses alternatives to specific population densities.~

   ~    Section l00.21(b) (2) prescribes a maximum population density at the time the application is approved and 40 years later.

If the population density exceeds the prescribed values, the Staff can still approve the site under Section 100.2l(b) (1) if the licensee demonstrates that there are no reasonably available alternative sites, or if the licensee offers "other considerations" why the site is nonetheless preferred.

WINSTON & STRAWN U.S. Nuclear Regulatory Commission March 24, 1993 Page 6 Alternatively, if the Staff concludes that it is not appropriate to generally revise the regulation to include an option that permits the use of plant design features where the exclusion area distance is less than 0.4 miles, existing sites that have an exclusion area distance less than 0.4 miles should be grandfathered from the requirements of proposed Section l00.2l(a) (l). For example, the proposed requirements in Section l00.2l(a) could be modified as follows (additional language is bolded): (a) Each reactor facility must have an exclusion area, as defined in§ 100.J(a) of this part. (l) For sites with a single reactor facility, other than reactor sites on which facilities have been licensed prior to (EFFECTIVE DATE OF THE FINAL RULE], the distance to the exclusion area boundary at any point ( as measured from the reactor center point) shall be at least 0.4 miles (640 meters). (2) For sites with multiple reactor facilities, other than reactor sites on which facilities have been licensed prior to [EFFECTIVE DATE OF THE FINAL RULE], consideration must be given to the following: If the reactors are independent to the extent that an accident in one reactor would not initiate an accident in another, the size of each exclusion area must be determined with respect to each reactor individually. The exclusion area for the site must then be taken as the plan overlay of the sum of the exclusion areas for each reactor * * *

• These changes would specifically exempt all existing reactor sites licensed prior to the effective date of the final rule from the requirement that the exclusion area be at least 0.4 miles.

Sincerely,

~tt?!:1-%/

Kathryn M. Kalowsky WINSTON & STRAWN

r--/00

            ~NIAGARA NUMOHAWK                                                                         '-' .

NIAGARA MOHAWK POWER CORPORATION/ 301 PLAINFIELD ROAD, SYRACUSE , N Y. 1321,2/TELEPHONE (3 15) 474-1511

• *93 MAR 26 A1 1 :28 HMPlL 0748 March 24, 1993 ' f Mr. Samuel J. Chilk U.S. Nuclear Regulatory Convnission Washington, DC 20555-0001

SUBJECT:

Proposed Rulemaking 10CFR Parts 50, 52, and 100, "Reactor Siting Criteria" (57 Federal Register 47802 - October 20, 1992 and 55601 - November 25, 1992)

Dear Mr. Chilk:

The purpose of this letter is to express Niagara Mohawk's concern on behalf of the nuclear industry on a world-wide basis that the subject proposed revisions will have a significant negative impact to both currently licensed and future plants without providing identifiable improvement to public health and safety. These negative impacts are well described and documented in comments provided by the Nuclear Management & Resources Council for the US nuclear industry, the Japanese Federation of Electric Power Companies for the Japanese nuclear industry, and others representing European viewpoints as well. We generally agree with the comments that are being made. For example, we agree that the adoption of the proposed criteria may adversely affect public perception regarding the acceptable safety of existing plant sites during their operating term and during plant license renewal proceedings. Therefore, we urge the NRC to reconsider their proposed changes. The nuclear industry has made steady progress, both in terms of performance and risk potential. Based on the accumulating reactor-years of good experience without detrimental environmental effects, public confidence has been building. This is the time to encourage this good experience with regulatory changes leading to further development and public benefits. If you would like to explore our thinking in more detail, please contact me. Sincerely,

                                                      ~

C. D. Terry Vice President, Nuclear Engineering

 /bp MJIY 111993 Acknowledged by card ..............................,...

Distribution: B. R. Sylvia J. R. Eichelberger R. J. Cazzolli J. Ronafalvy (NUMARC) T. Imai (JAPC/FERC) Y. Nagai (GPUN) Records Management File

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Buyers Up

• Congress Watch
• Critical Mass
• Health Research Group
• Litigation ~ upMAR 25 p 3 :Q7 Ralph Nader, Founder UNITED STATES NUCLEAR REGULATORY COMMISSION 10 CFR PARTS 50, 52, AND 100 Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of petition for Rulemaking From Fre e

- Environment, Inc. et al. 57 F.R. 47802 COMMENTS OF PUBLIC CITIZEN'S CRITICAL MASS ENERGY PROJECT INTRODUCTION The Nuclear regulatory Commission has issued for comment a proposed rule and a denial of a petition for rulemaking regarding reactor site criteria. Integral to the discussion is the Commissions change in regulatory philosophy of defense-in-depth and a retreat from the use of accident consequences known as the source term in regulation. The Commission staff claims that this change in regulation is necessitated by the fact that dose considerations effect plant design more than siting. (Staff Study on Source term Up date and decoupling siting from design, SECY-90-3 4 1, October 4, 1990, p. 4.) In reality this proposed rule is the realization of a long sought goal of the nuclear industry to remove the consideration of accidents from the siting of nuclear power plants. I. THE U.S. NUCLEAR REGULATORY COMMISSION FAILS TO PROVIDE A REASONED ANALYSIS FOR CHANGING THE PRESENT RULE AS REQUIRED UNDER THE ADMINISTRATIVE PROCEDURE ACT Under the Administrative Procedure Act, an agency that rescinds a rule is required to supply a reasoned analysis for the change beyond that which may be required when an agency does not act in the first instance. (Motor Vehicle Manufacturers Assn v. State Farm Mutual Automobile Insurance Co., 463 U.S. 29, 42 [1983]) . In the present instance, the NRC's rationale fails to provide ample justification for the proposed rule change. MAY t 11993 Acknowledged by card .................~:............" 215 Pennsylvania Avenue SE

• Washington , D.C. 20003 * (202) 546-4996
• FAX: (202) 547-7392 e~" @ Primed on Recycled Paper

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The NRC claims that its experience with siting has led to the conclusion that source term considerations effect plant design more than siting. According to the NRC, "decoupling would replace existing site dose calculation requirements (which traditionally have affected plant design more than siting) with explicit requirements more directly related to acceptable site characteristics." (Staff Study on Source Term Update and De-coupling Siting form Design, SECY-90-341, Oct.4, 1990, p. 4). The Commission gives little evidence to support this position but rather bases its supposition upon agency experience. While the Commission certainly has experience siting reactors, one could legitimately argue that this experience has not been translated into expertise. Public Citizen believes that the reason the source term has effected plant design more than siting is the fact that the NRC sited nuclear reactors and then attempted to justify the risk posed to the public. The licensing battle around Limerick Unit 2 is i ns tructive. Having already sited the facility the Commission then had to address plant design features in order to justify its operat ion. The problem lies not with the source term but with the Commission's having under evalua ted the importance of approp riate siting of a nuclear power plant. A mere parenthetical allusion to the need for altering the present rule hardly constitutes a reasoned analysis as contemplated by the administrative procedure act. II. THE NRC SHOULD NOT REMOVE THE SOURCE TERM FROM REGULATION The Nuclear Regulatory Commission should not allow the removal of source term considerations from regulation. In fact, in the absence of a coherent safety goal policy, the site dose calculations provide a benchmark against which to measure the appropriateness of a reactor site. The NRC's desire to rid regulation of accident dose considerations is quite understandable. The NRC and the nuclear industry could not justify nuclear power plant operation if the source term were updated rather than eliminated. Nuclear industry efforts in the 1970's and 1980's concentrated on reducing the source term in order to persuade the public that nuclear power was perfectly benign. The NRC's risk studies rather than assuaging the public's fear of nuclear power has actually fanned it. NUREG-1150 completely undermines the assumptions necessary for the source term calculation. Basically, it explodes the myth that during a severe accident the reactor containment will hold. In its original form NUREG 1150 concluded that early containment failure could not be ruled out in a severe accident for any of the containments studied. (Reactor Risk Reference Document, NUREG-1150, February 1987, p. ES-14). If

we we r e t o create ex c lusion zones and low population zones bas e d upon the reality of early containment failure, the public would be too alarmed to ever allow another nuclear reactor to be constructed. Additionally, the NRC wishes to remove dose calculations from decisions pertaining to engineered safety features (ESF). Again the NRC's rational for this change is lacking a well reasoned analysis. The staff states that "(t)he ESF requirements will be based on best engineering judgement and will not resort to dose calculations." (SECY-90-341 at p. 12). Thus, the NRC is moving from a quantifiable standard against which safety may be judged to a standard which is subjectively based upon the "judgement" of the agency's and industry's engineers. This does not constitute a sound basis for regulation. The staff then participates in the fantasy that "if the development of these ESF requirements proves intractable, dose criteria and reference to the new source term can be added to Part 50." (SECY-90-341 at p. 12). Public Citizen bel i eves that if the NRC successfully r emoves the source term from regulation, dose considerations will never again be used in regulation. Basically, the industry would never allow the reintroduction of a regulatory standard which they could not meet. CONCLUSION The Corranission asks the public to address several questions in regards to the proposed rule. Public Citizen will gladly address these questions however, they do not go to the heart of the Corranission's proposal. Wrapped up in the technical questions are those which are more basic. Does removing the source term from regulations enhance public safety? This rule does not protect the public; it does, however, protect the nuclear industry's ability to site a nuclear reactor. Public Citizen believes that the NRC should update the source term and acknowledge what the public already knows--that the myth of containment is a lie. One need only examine the history of the General Electric's Mark I design and the fallacy of pressure suppression containment to realize the extent to which the NRC is participating in a fantasy. The NRC's proposed rule is not regulation; it is capitulation.

QUESTIONS A. Reactor Siting Criteria (Nonseismic)

1. Should the Commission grandfather existing reactor sites having an exclusion area distance less than 0.4 miles (640 meters) for the possible placement of additional units , if those sites are found suitable from safety considerations?

As noted in the regulatory analysis accompanying the proposed rule, "the effect of these requirements is to set both individual and, to some extent, societal limits on dose (and implicitly risk) . . . . " This being the case, the grandfa thering of exi s ting reactors which violate the .4 mile exclusion zone would deprive certain individuals of equal protection under NRC regulations. The NRC should not grandfather those reactor sites which violate the .4 mile exclusion zone requirement. Ideally, the NRC should look to phase out those reactors which over time have come to present a greater risk to the public health and safety. Since an NRC required phase-out is unlikely, the NRC should compensate by requiring enhanced emergency planning procedures for those closest the reactor.

2. Should the exclusion area distance be smaller than 0.4 miles (640 meters) for plants having reactor power levels signi-ficantly less than 3800 Megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example, 0.25 miles or 400 meters)?

The NRC should not allow for smaller exclusion areas for smaller reactors. The result would allow reactors in more densely populated areas which seems antithetical to the purpose of have an exclusion zone in the first place. It would require a case by case examination of the reactor/site interface which under the new reactor licensing requirements is impossible. The application for an early site permit in the "one step licensing" scheme does not envision the size of the reactor to be placed on that site.

3. The Commission proposes to codify the population density guidelines in Regulatory Guide 4.7 which states that the population density should not exceed 500 people per square mile out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter .

A. Should numerical values for population density appear in regulation or should the regulation provide merely general guidance, with numerical values provided in a regulatory guide . The NRC should include numerical values for population density in the regulation. To place the values in a regulatory

guide would essentially remove the teeth of the regulation. If its in the regulations it is, at least hypothetically, enforceable. B. Assuming numerical values are to be codified, are the values of 500 people per square mile out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter appropriate? If not , what other numerical values should be codified and what is the basis for these values? As a public policy consideration, it would seem the NRC would want to site reactors as far from population centers as possible. One way to accomplish this would be to decrease the allowable population density. While Public Citizen has no specific values it would like to see codified, the values adopted by NRC should reflect certain realities. The values should acknowledge the reality of the Chernobyl accident and the fact that early containment failure can not be rule out with high confidence for any of the plants studied in the Reactor Risk Reference Document, NUREG-1150. C. Should population density criteria be specified out to a distance other than 30 miles (SO km), for example, 20 miles (32 km)? If a different distance is recommended, what is the basis . The population density criteria should be specified out to a distance of at least 30 miles. A case could be made to extend this distance based upon the experience of Chernobyl and the likelihood of early containment failure in the event of a severe accident.

4. Should the Commission approve sites that exceed the proposed population values of 10 CFR 100.21, and if so, under what conditions?

Siting is the most basic precaution the commission can take to ensure the public health and safety. If the NRC is going to spend the tax payers money to rewrite regulations they should not simultaneously create loop holes in those regulations. What's the point of having regulations if the NRC is not going to enforce them.

5. Should holders of early site permits, construction permits and operating license permits be required to periodically report changes in potential off site hazards ( for example, every 5 years within 5 miles)? If so what regulatory purpose would such reporting requirements serve?

Periodic reporting of changes in the potential for off-site hazards is necessary due to the allowance of early site permits and site banking under part 52. Again this type of review of site could preclude the protracted confrontations experienced at Seabrook and Shoreham.

6. What continuing regulatory significance should the safety requirements in 10 CFR Part 100 have after granting the initial operating license or combined operating license under 10 cfr part 52?

The NRC should use the safety requirements in 10 CFR Part 100 as a bench mark to measure plant safety in the absence of a coherent safety goal.

7. Are there certain site meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are the conditions that can not adequately compensated for by design features?

As noted in the discussion of the ACRS, there are meteorological conditions that should preclude siting of a nuclear power plant. As an example areas which experience long - term inversions should not be considered as amenable sites. March 24, 1993 Respectfully Submitt~

                             ~ ~cio, Esq.

Ene r gy Campaigner Critical Mass Energy Project

                         'DOCKET NUMBER            R p 52 PROPOCt:D EULE

(>1 FR L/7~0;J..) Westinghouse Energy Systems Box 355 Electric Corporation *93 HA 25 P1 :11 Pittsburgh Pennsylvania 15230-0355 ET-NRC93-3846

                                                     ,_ f '                               NSRA-APSL-93-0092 t    l,f\L March 24, 1993 Mr. Samuel J. Chilk Secretary, Office of the Secretary of the Commission U.S. Nuclear Regulatory Commission Mail Stop 16 G15 Washington, D.C. 20555

SUBJECT:

Proposed Rulemaking 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria" (57 Fed Reg. 47802, October 20, 1992 and 57 Fed Reg. 55601, November 25, 1992)

REFERENCE:

Letter, William H. Rasin (NUMARC) to Samuel J. Chilk (NRC), dated March 24, 1993

Dear Mr. Chilk:

The purpose of this letter is to provide Westinghouse comments on the subject proposed rulemaking. This rulemaking would codify population density and minimum exclusion area size numeric criteria for siting future nuclear power plants. The proposed rulemaking also revises seismic and earthquake engineering criteria for future nuclear power plants. As a worldwide supplier of over 100 nuclear power plants and as an applicant for Design Certification of an advanced nuclear plant design under 10 CFR Part 52, Westinghouse has been actively involved in the industry review of the proposed rulemaking. We have participated in the preparation of industry comments on the proposed rulemaking and we fully endorse the comments submitted on this subject by NUMARC in the reference letter. Westinghouse shares the industry concerns on this subject as expressed in the NUMARC letter. The following comments are offered to provide further amplification on two of the main areas of concern. As stated in the Federal Register notice, one of the USNRC objectives in proposing this rulemaking is to "[s]tate the criteria for future sites that, based upon experience and importance to risk, have been shown as key to protecting public health and safety." The proposed rulemaking specifies population density and minimum exclusion area siting criteria, however, experience has shown that the key criteria for siting have been the radiological consequence evaluation factors contained in the current 10 CFR Part 100. The current criteria have been used to safely site all licensed nuclear power plants in the United States. The application of different criteria for new plants is inconsistent with the stated objective and has the potential for significant unintended impacts on public perception regarding the acceptable safety of both existing nuclear power plants as well as advanced light water reactor designs. Adoption of the proposed population density/exclusion area siting criteria would send an inappropriate message on the acceptability of current licensed nuclear power plant sites as well as the siting of U.S. advanced reactor designs, both in the U.S. and worldwide. The radiological dose consequence evaluation factors in the current 10 CFR Part 100 are the key criteria for future sites and these criteria should be maintained in the rule. II& 1 J fQr:17 0874A Acknowledged by card ..........................~=**

ET-NRC93-3846 March 24, 1993 NSRA-APSL-93-0092 A second stated objective in proposing this rulemaking is to "[p]rovide a stable regulatory basis for seismic and geologic siting and applicable earthquake engineering design of future nuclear power plants ... " The proposed rule includes a requirement for both deterministic and probabilistic evaluations. This requirement should be removed. The requirement for both types of evaluations without a clear technical procedure for reconciliation of differences will not provide the "stable regulatory basis" sought by the proposed rule. We urge the NRC to give serious consideration to the alternative seismic siting decision process developed by the industry and outlined in the NUMARC letter. We appreciate this opportunity to comment on the proposed rulemaking. The issues being addressed by the proposed rulemaking are of significant importance to the nuclear industry and Westinghouse urges the Commission to carefully consider the above comments as well as the industry comments provided in the NUMARC letter. Very truly yours, a~//~ N. J. Liparulo, Manager Nuclear Safety and Regulatory Activities /nja cc: Mr. J. Taylor, Executive Director of Operations, NRC Dr. T. Murley, Director, Office of Nuclear Reactor Regulation Mr. E. Beckjord, Director, Office of Nuclear Reactor Research Mr. W. Rasin, Vice President & Director, NUMARC Mr. B. McIntyre, Westinghouse 0874A

MAR 24 1993

                                                                             *93 MAR 25 P1 :11 Secretary U.S. Nuclear Regulatory Commission Washington D.C. 20555

• I y r vI f Attention: Docketing and Service Branch Subj e ct: Transmittal of Comments from a U.S. Department of Energy {DOE) Review of the U.S. Nuclear Regulatory Commission's {NRC) Proposed Revisions to 10 CFR Parts 100, 50, 52, and Related Regulatory Guides This letter transmits comments from a DOE review by the Office of Civilian Radioactive Waste Management {OCRWM) of NRC's proposed revisions to 10 CFR Parts 100, 50, 52, and Regulatory Guides.

The scope of the proposed rule in 10 CFR Parts 50, 52, and 100 clearly indicates that the proposed revisions are applicable to power and test reactors only. NRC, through NUREG-1451 Staff Technical position {STP) on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Repository, has explicitly stated that Appendix A of 10 CFR Part 100 does not apply to the geologic repository program. DOE would like to emphasize the inapplicability of guidance in Appendix B of 10 CFR Part 100 to the geologic repository program. Given the apparent long-term recurrence rates and very low slip rates at Yucca Mountain, the NRC's acceptance of a probabilistic methodology as an integral part of the seismic hazard analysis is viewed very favorably by OCRWM. The incorporation of probabilistic analysis in the proposed revisions to Part 100 represents a positive step, allowing for the examination of uncertainty and the consideration of alternate tectonic models. However, the manner in which probabilistic and deterministic methods are to be combined or weighed is unclear, and needs more specific definition in the Regulatory Guides. Regulatory Guide DG-1015 calls for geologic reconnaissance and literature reviews to cover a radius of 320 km from the site to identify seismogenic sources. There is no rationale for this large radius. It may be more appropriate to first conduct sensitivity studies to determine to what distance from the site a seis mogenic source may be significant. MAY 111993 Acknowledged by card .......................::::::;

Our specific connnents on the proposed revisions are enclosed. Should you have any questions, please contact Chris Einberg of my office at (202) 586-8869. Sincerely, OJ~

                     /2!~ate        Director for Systems and Compliance Office of Civilian Radioactive Waste Management Enclosure As Stated cc w/enclosure:

C. Gertz, YMPO T. J. Hickey, Nevada Legislative Committee R. Loux, State of Nevada D. Bechtel, Las Vegas, NV Bureka County, NV Lander County, Battle Mountain, NV P. Niedzielski-Eichner, Nye County, NV

w. Offutt, Nye County, NV C. Schank, Churchill County, NV F. Mariani, White Pine County, NV V. Poe, Mineral County, NV J. Pitts, Lincoln County, NV J. Hayes, Esmeralda County, NV B. Mettam, Inyo County, CA C. Abrams, NRC J. Holonich, NRC

REVIEW COMMENTS ON NRC's PROPOSED CHANGES TO 10 CFR SO, 52, AND 100, AND SUPPORTING DOCUMENTATION GENERAL OBSERVATIONS The comments provided below are not directly applicable to the DOE repository program or MRS program. However, DOE recognizes that 10 CFR 72, which governs MRS siting, does invoke Appendix A of 10 CFR 100 for sites west of the Rocky Mountain front. The proposed revisions to the regulations are not applicable to the evaluation of site suitability at the proposed Yucca Mountain repository site. However, the explicit recognition and requirement to use the probabilistic seismic hazard approach as part of the seismic hazard evaluation for determining the Safe Shutdown Earthquake (SSE) for a nuclear power plant, is a clear indication that the NRC is viewing the probabilistic approach as an integral part of any seismic hazard analysis. The in001p0ration of probabilistic analysis represents a positive step, allowing for the examination of uncertainty and the consideration of alternate tectonic models. The following comments relate to certain parts of the proposed Rule for 10 CFR 100; Appendix B, Appendix S, Regulatory Guides DG-1015, 1016, and 4003; and to 10 CFR 50. PROPOSED RULE 10 CFR 100 Section 100.20 Section 100.20 (c)(3) includes ground water velocity and the distance to surface-water bodies as factors to be obtained from on-site measurements. Because it is not directly measurable, ground-water velocity should not be regarded as a discriminating site factor. The configuration of the potentiometric surface and the hydrologic properties of the media through which ground-water moves are the more basic and relevant site factors, which control ground water velocity. A site factor at least as important as the distance to surface-water bodies is the depth to the water table beneath the site under both present and possible future conditions. Appendix B Appendix B, Sections II and IV state that both detenninistic and probabilistic evaluations must be conducted for seismic and geologic design bases. Appendix B does not address what will be done if the two approaches (deterministic and probabilistic) come up with substantially different results. An example is where the deterministic earthquakes have extremely long return periods. A non-uniform level of hazard is inherent in all deterministic approaches. One way around this basic problem is for the NRC to clearly March 23, 1993 Enclosure

establish, by definition, a mioimnm goal for acceptable design, and use the probabilistic and deterministic information to document that the goal has been reached. The NRC should emphasize achieving the goal, rather than demonstrating how the goal was achieved. For deterministic evaluations, this means that the estimated maximum credible earthquake (MCE) will have to be accompanied by an estimated recurrence interval. The relative roles of detemrinistic and probabilistic methods, while appropriate for the seismically active western United States, seem less appropriate for the stable continental region (SCR) of eastern North America. For much of the SCR, detennioistic earthquakes can not be related to known faults, but rather are attributed to source zones (seismogenic sources). The cause of earthquakes in these areas is very uncertain. In this situation, the probabilistic analysis should be given more weight in assessing a Safe Shutdown Earthquake. Where seismogenic sources can only be defined as zones, the probabilistic approach should be given more weight. In Appendix B, Section ID (1), the definition of "capable tectonic source" follows the same form as the definition in Appendix A. This requires detennioiog the number of events that have occurred over the last 500,000 years. Clearly separating and dating all the individual faulting events over such a time period can be extremely difficult to do. Most often the :resuhs are unclear. How does the NRC intend to address the uncertainty here? Most commonly one or more events may be recognized to have occurred within 500,000 years. Also in Appendix B, Section m (1), this definition includes faults that have had displacement once in the past 50,000 years, instead of once in the past 35,000 years as in Appendix A. This change presumably reflects a demonstrated need for greater conservatism providing an increased level of safety and health. However, the Appendix does not state that a change (to 50,000 years) would provide such an increased level of safety and health. Extending the time period effectively adds additional and perhaps unnecessary conservatism to the seismic hazard evaluation. In addition, if the extended time is because of improved age dating techniques, 50,000 years is not a particularly good age. Current dating techniques such as carbon 14 are not particularly reliable at 50,000 years. REGULATORY GUIDE DG-1015 Use of Probabilistic versus Determinmie The NRC requested comments on SPECIFIC ISSUES, which were stated in an attachment to DG-1015. Issue #1: How should the use of deterministic and probabilistic evaluations be combined? Comment: The draft regulations and guidance reflect the current state of the practice in advocating the use of both deterministic and probabilistic methods. It is uncertain why the staff is proposing two March 24, 1993

parallel evaluations rather than using a serial approach. Deterministic evaluations generally provide the foundation for probabilistic evaluations. For example, seismic source zones would have to be identified from a deterministic analysis. The same zones should then be used in the probabilistic analysis that builds upon the information developed from the deterministic approach. Rather than weighing the two approaches, the probabilistic analysis has the merit of being able to explicitly examine uncertainties in a deterministic approach with regard to alternate tectonic models, the choice of maximum magnitude earthquakes, and the choice of an attenuation relationship. Issue #3: Should the median values of the seismic hazard analysis be used in determining controlling earthquakes? Comment: It has been common practice to use the 85th percentile to provide a more conservative estimate. The staff should provide the rationale for why they believe the median would provide a more suitable estimate. For example, would the median be more suitable with regard to less significant changes to the earthquake database? Numerical values for SSE Annual Probability of EJ:rffi'Hnce In Appendix B to 10 CFR 100 (section V), the NRC has set the acceptable level for estahUsbing the SSE based on comparison to the median annual probability computed for the current population of nuclear power reactors. This is a vague requirement that leaves substantial room for interpretation and argument. It is recognized that Appendix B of Draft Regulatory Guide DG-1015 provides definitive numerical values for this goal. The NRC should continue to provide actual numbers for the SSE annual probability of exceedance goal - if not the 10 CFR Part 100, at least in the subtler documents. Probabilistic Analyses in Eastern USA and Western USA In pages 6 through 8 of DG-1015, the discussion divides the U.S. into eastern and western sections based in part on the differences in observed relationships between seismicity and tectonic features, and in part on the availability (in the eastern section) of accepted probabilistic analysis done by EPRI and LLNL. This discussion does not appear to fully recognize the quality and value of a well-based and well-documented probabilistic analysis for the western U.S. For western sites, the examples given use only a deterministic approach to determine controlling earthquakes. DG-1015 should include examples of acceptable probabilistic analyses for western sites. To accommodate the different relative strengths of probabilistic and deterministic analyses for different areas, NRC should explicitly allow the applicant to provide its own defensible method for combining the two analyses, rather than follow a prescribed procedure. The use of a value for the annual probability of exceeding the safe shutdown earthquake ground March 23, 1993

motion based on current population of power plants would not be useful for regional probabilistic analyses in the western U.S., where there are not enough licensed power plants to be statistically meaningful. A set figure, such as 10"4/ year (based on the LLNL analysis) may be preferable. Location of Seismogenic Source Area Earthguake On page 9, Section B(4.) (1) of DG-1015, it is stated that for the seismogenic source area that incorporates the site, the source earthquake for design purposes is assumed to occur 15 km from the site. This distance constitutes new guidance. The rationale for the choice of 15 km should be provided in DG-1015. Also, guidance should be provided for those cases where the seismogenic source is closer than 15 km. Controllio& Earthguakes Pages 9 and 10, Section B(4.): For identifying controlling earthquakes in the event of significant differences between deterministic and probabilistic analyses, guidance in DG-1015 is unclear on how to choose the design basis earthquakes, and whether one analysis or the other should prevail. If the controlling earthquakes are not easily identifiable, guidance is again insufficient. Additional guidance is needed to demonstrate acceptable conservative approaches. Use of LLNL, EPRI or Both Methods The NRC should make a decision on the proper method to use. It is preferred that both methods be used, and that they be used to represent the bounds for the seismic hazard for a given site. Specific Issue# 2 (for DG-1015) requested for Comment by the NRC is paraphrased as follows: Issue #2: Is the proposed procedure in Appendix C of DG-1015 adequate to determine controlling earthquakes from the probabilistic analysis? Comments: The proposed procedure appears to assume prior knowledge of the LLNL and EPRI methods as well as the current state of the practice. The use of the example for the Ea.stem U.S. site_is __ _ a gooo idea, but the example may have to be more of a tutorial to be more generally useful. The guidance provided in the regulatory guide text and in Appendix C appears to be too prescriptive. While it is encouraging to note the regulatory flexibility contained within the proposed regulations and guidance, the NRC may want to consider means for encouraging flexibility in the applicants' approach(es) to determining controlling earthquakes. National Standard for Area of Search and Seismic Sources DG-1015, Section C, page 11, states that "regional investigations such as geological reconnaissances and literature reviews should be conducted within a radius of 320 km (200 mi) of the site to identify seismic sources". What significance would a seismogenic source have on March 23, 1993

a facility 150 or 200 miles away? Perhaps a better approach would be to first conduct sensitivity studies to better determine what distance is "significant". Because the contnlrution of a source to haz.ard at a site depends on the source size, distance from the site, ground motion attenuation properties, and site effects, application of the static standard across the United States seems inappropriate. While it may be adequate in one region, it could be overly conservative in another. The derivation and significance of the 320 km radius should be discussed/justified in the regulatory guide. REGULATORY GUIDE DG-1016 On page 3, section C. , the second paragraph states that seismic monitoring instrumentation should be operable at all times. This is too restrictive. To ensure operability at all times may require providing completely redundant instrumentation throughout the plant. A possible alternative to the NRC's requirement would be to place technical specification restrictions on plant operation after a certain specified time period when seismic instrumentation is out of service. REGULATORY GUIDE DG-4003 One page 26, section 7, the NRC has proposed that design basis events resulting from accidents at industrial, military, or transportation facilities in the vicinity of the plant have a probability less than 1 x 10-7 per year. This value is too low. Other events considered in the design basis, except tornado, have a cutoff probability much higher than this. The value suggested here is too restrictive/conservative. On page 22, subsection 4.4, second bullet: replace absotption with adsorption. On page 22, subsection 4.4, third bullet: "ground-water velocity" is inappropriate because it is an ambiguous concept. Also, the depth to the water table (more specifically, the water-table configuration beneath the site) should be required information. APPENDIX S TO 10 CFR so The second paragraph on page 47815 states that "the proposed regulation would move the location of the seismic input motion control point from the foundation-level to free-field, at the free ground surface or hypothetical rock outcrop, as appropriate." the reason for use of the phrase "as appropriate" is not intuitively obvious to the reader. Page 2.2.2-8 of Proposed Revision 3 to Standard Review Plan 2.5.2 provides a good explanation of when the location of the seismic input motion control point should be specified at the ground surface, and when it should be applied at the rock outcrop. A brief clarification of which location is appropriate should be included in the Appendix S paragraph. March 23, 1993

noc~<ET lUMBER Pl 5 ?Y~

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                     *** CORRECTION TO***

COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC. ("OCRE") ON PROPOSED RULE, "REACTOR SITE CRITERIA; INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PETITION FOR RULEMAKING FROM FREE AMERICA, INC. ET AL.", 57 FED. REG. 47802 (OCTOBER 20, 1992) OCRE is hereby amending its comments, dated March 22, 1993, in the above-captioned rulemaking, to correct the following typo-graphical error: On page 11, first paragraph under C. Use of Probabilistic vs. Deterministic Methods, 14th line of that paragraph, "lOE-4" should be "lE-4". Respectfully submitted, Susan L. Hiatt Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 (216) 255-3158

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l oocKET NUMBER PROPOSED RULE Pl 5.?:J,, 5 2-- J-l O 0 C7" r- rt '11 ro ;i.) March 24, 1993 Secretary of the Commission Docketing and Service Branch U.S. Nuclear Regulatory Commission Washington, o.c. 20055 Comments On Proposed Rule for Reactor Site Criteria Mr. Secretary: Nuclear Information and Resource Service (NIRS) is very concerned about the new rules the Nuclear Regulatory Commission is proposing on reactor siting criteria for nuclear power plants as published in the Federal Register 47802-47821, October 20, 1992. It is apparent that the intent of the new rules is to facilitate electric utilities in finding sites for a new generation of nuclear power plants by allowing reactors to be sited in more densely populated areas. The rule also ends the current linkage between source term calculations and reactor siting. Furthermore, NRC proposes to "grandfather" existing operational nuclear power plants from proposed minimal exclusion zone regulations with this exemption possibly being extended to new reactors built on existing sites. NIRS takes the position that these and other asP,ects of the new rule on reactor siting criteria further marginalize public health and safety in favor of easing the licensing and regulatory climate for the nuclear power industry. NIRS takes the position that this violates the fundamental objective of the NRC. I.REACTOR SITING CRITERIA (NONSEISMIC) A. 11 Decoupling 11 reactor siting from accident radiation dose calculations NRC claims that numerous risk studies ori radioactive material releases to the environment under severe accident conditions have all confirmed that the present siting practice is expected to effectively limit risk to the public, principally citing the 15 years of service to the 100% ,acyc/ed paper grassroots environmental movement

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"Reactor Safety study" (WASH-1400) and "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants" (NUREG-1150). The proposed criteria ¥fOUld then "decouple" siting from source term and dose calculations to the public arguing that these factors relate to reactor design rather than reactor siting. - First, NIRS believes that NRC "has put the cart before the horse by proposing the decoupling of, source term from siting criteria in absence of a final design approval for advanced reactor designs. In fact, the NRC certification process continues to fall behind schedule stemming from a number of unresolved design issues. _ NIRS is not convinced that the new reactor technology represents a "failsafe $ystem" which decoupling would suggest. source term and dose calculations regulations were intended to help mitigate the consequences to, the public and environment from a nuclear reactor accident. Source term information provides the essential link in estimating what the impact on a particular geographical area around the plant after any given initiating event { such as a pipe break or an ECCS actuation signal failure.) Geographic location and associated demographic therefore remain i~portant factors associated with the type and design of power station being proposed. It is illogical for NRC to assume that increasing the number of nuclear power plants is any reason to move towards less conservative regulations for siting. NIRS objects to NRC assistance to a nuclear industry public relations campaign to sell the public on "inherently safe reactor designs" for what must be vigilantly recognized as an inherently dangerous technology. Decoupling source term from reactor siting is, in fact, tantamount to abandoning concern for public health and safety to accommodate early s-i te regulations. NIRS argues that decoupling source term from siting criteria also undermines emergency planning regulations for present and future populations. NIRS argues that current NRC siting criteria and associated accident consequence analysis need to incorporate more conservative margins of safety, not less. NIRS argues that source term calculations should remain part of the siting criteria and_in fact should be expanded to compensate for.uncertainties, errors and non-conservatisms in the overall analysis of the entire spectrum of possible accidents. For example, one source term non-conservatism under current siting criteria recognizes only iodine as the surrogate for all non-gaseous fission products. This needs to be compensated for. Another non-conservatism in the present use of source-term regards exposure to "man equivalent" only. Exposure to fetuses, infants, children, -the elderly, and people with chronic illness are not included in current source term and population exposure estimates. Similarly, it is non-conservative to consider only radiation exposure and disregard the synergistic effect of

simultaneous exposure to other man-made toxic hazards and associated environmental degradation (i.e. ozone depletion.) NIRS argues that source term calculation should remain part of the siting criteria as an effort toward upgrading to more conservative estimates for further Probabilistic Risk Assessment (PRA) studies in more accurately estimating the likelihood, progression, and consequence of various potential severe nuclear power plant accidents. Population density numbers developed for determining the risk area significantly figure into such studies:- B. Exclusion z~ne of

• 4 miles establishe.d for new reactor sites and th-a 11 grandfathering 11 of old reactor sites that do not meet stricter standard.

NIRS recognizes that the current exclusion zone practice is inadequate for providing the public with any high degree of protection from radioactive releases resulting from routine operation or operation of accident mitigation,,systems. The areas arbitrarily established for the current reac'tor population have no numerical size requirement. The present regulation's assessment of the consequences of a radioactive release were undermined by shortcomings in the PRA process including; 1.) failure to account for aging of reactor structures and components; 2.) failure to account for licensee's operation of reactors in violation of their technical specifications with and without NRC approval; 3.) failure to account for partial system failures; 4.) failure*to account for uncertainties regarding containment weaknesses NIRS approves of establishing stricter minimum requirements for exclusionary zones. However, NIRS questions whether NRC is merely establishing yet another arbitrary utility control zone under the new standard. This concern is compounded by the proposed elimination of the use of a postulated* source term, assumptions regarding mitigation systems and dispersion factors, and the calculation of radiological consequences to determine the size of the zone. Rather these factors and an upgraded PRA process should be incorporated to determine what would be the size of an effective exclusion zone. For the same above stated reasons, NIRS opposes reduced exclusions zone distances for smaller power reactors. NIRS objects to the "grandfathering" of the 23 existing sites that could not meet the. proposed standardized exclusionary zone. NRC continues to portray the operation of nuclear power plants as a benign technology, as if we were being asked to consider grandfathering an outhouse within city limits. If NRC is going to

formulate standards, the basis for said standard should have solid \ foundations and it is then expected that NRC enforce the regulations at the substandard sites. "Grandfathering" of aging and increasing more decrepit nuclear power plants underscores the NIRS'concern that the proposed standard represents "old wine in a new skin." For the same reasons, NIRS objects to the siting of new reactors at "grandfathered" sites. The public trust is further damaged by NRC formulating willynilly standards supposedly based on a public health and safety objective. New reactors should never be built where the sites are considered to be substandard. ( {

c. Higher population density figures for the low population zone NIRS is opposed to proposed NRC rule changes on population density and the NRC failure to* consider population restrictions beyond a 30 mile radius. NIRS takes the position that population d~nsity for reactor siting criteria should not be increased; it should be decreased.

The 1979 Siting Task Force held that from the exclusion ~one-to 5 miles the maximum population density should be at most 100 people per square mile; from 5-10 miles, 150 people per square mile; and from 10-20 mile, 400 people per square mile., NRC justifications for increased population density figu~es in ~e low population zone are based in the Commission's Policy Statement on Safety Goals quantitative health objective in regard to estimates for latent cancer fatalities and Jand contamination. NRC analyses that "population density restrictions out to 40 miles-could make it difficult to obtain suitable reactor sites in some regions of the country" is an outrageous admission on the part of NRC that easing of reactor siting criteria is more a priority than public heal th and safety. It can be construed. that in this case "suitable reactor sites" has more to do with marketability of electricity than with public safety. In light of the far-reaching consequences demonstrated in the Chernobyl accident, the public is likely to be unwilling to believe that radiation contamination can be limited to arbitrarily drawn political lines, such as the 10 mile Emergency Planning Zone. While NIRS and the public are willing to distinguish technical design differences between the RBMK reactor and US models, both operational and new design, it is now broadly recognized that the release of any fission reactor's* radioactive inventory once borne on the weather knows no arbitrary established boundary. \ NIRS objects to NRC basing any of it's regulations on the marketability of nuclear power and rea$serts that protecting the

public health and safety is the NRC primary respons.ibility in regulating nuclear power. NIRS takes the position that population restriction zones should be extended out to the currently established accident interdiction limits outlined in the 50 mile ingestion pathway zone (IPZ). NIRS would argue that holders of early site permits and* construction and operation licenses be required to periodically report on demographic changes within the established IPZ. License holders found to exceed established population values would loss their license by that time. NIRS woµId argue that any nuclear power plants that are projected to break the 1,000 persons/sq.mi. limit at 30 miles by the year 2000 be Shut down. This reactor population would include the Limerick, Indian Point, Fermi 2, and Haddain Neck stations. - C NIRS urges ~hat additional action - be taken to establish fUrther restrictions on nuclear power plant siting in prime agricultural and food production areas within the IPZ. A numerical value based on food productivity along with population density would be assigned to protect this vital natural and economic resource from the risk of land contamination arising from the release of cumulative long-lived :radioactive materials in , the course of routine operation, licensee operation outside_ of technical specifications, partial systems failures, and in the event of a severe reactor accident. D. Minimum standoff distances NRC proposed regulations on evaluation of man-related hazards and site acceptability do not go far enough. NIRS concurs with the 1979 siting Task Force recommendations to establish minimum standoff' distances for all nuclear power plant sites from major airports and military air bases, Liquid Natural Gas terminals, large propane and natural gas pipelines, explosive and toxic material industr:lal sites, major dams, and capable faults. NRC is g.eferring its duty to protect public health and safety by failing to incorporate

• tough minimum standoff distance limits in the siting criteria.

Holders of early site permits, construction and operation licenses should be required to periodically report changes in potential man-related or other offsi te hazards. Nuclear power stations identified with potential offsite hazards within established

• standoff z*ones would then enter a license review process to mitigate the hazard or lose t~eir license.

NIRS proposes that NRC establish a minimum standoff distance

from coastlines prone to hurricanes. In August, 1992 Turkey Point Units 1 and 2 were subject to a direct hit of Hurricane Andrew. The reactor operators were caught off guard by the early arrival of the hurricane and the reactors were in hot standby at the onset of the storm. Among the damage reports, NIRS notes that the reactors lost all fire suppression capability, radiological monitoring, site security systems, sustained a 3 hour control room communication blackout, and lost all offsite emergency notification systems for several weeks. The emergency planning zone was in chaos for months following the disaster. The failure of plant operators to accurately predict the arrival of the storm so as to achieve cold shutdown, the failure of so many vital plant systems relevant to accident mitigation, and total collapse of the emergency planning zone focuses concern that there are certain meteorological conditions,that preclude the siting of nuclear power plants. Severe coastal storms adversely affecting nuclear power plant operation is in evidence since Hurricane Andrew. The Pilgrim station experienced a reactor scram on December 12, 1992 as a result of storm deposited salt spray which caused a 345 KV switchyard flashover. Following a weekend coastal storm on March 17, 1993, both Crystal River and Brunswick nuclear power plants experienced loss of offsi te power due to degraded swi tchyard equipment from ocean salt spray deposits. The plants entered into unusual event status when emergency diesel generators were used during a loss of offsite, power to clean electrical insulators. Extended severe weather conditions could create such degraded conditions to stress emergency systems and hinder maintenance operations at added risk to safe operation. Consequently, NIRS argues this to be_ further evidence for the need to incorporate coastlines into a minimum standoff distance criteria.

      *E. state authority and nuclear power plant siting.

Contrary to the recommendation of its 1979 task force, the NRC's proposed criteria would not recognize final state denials of siting permits, even, apparently, when the denial is based on factors other than: radiolo'gical safety and are clearly within the state's authority. This would be an unacceptable expansion of federal pre-emption and undoubtedly would lead to substantial litigation, thus defeating any purpose of regulatory stability. / NRC has argued that "this recommendation would give a* State the authority to grant issuance of a construction permit for a nuclear facility. Only the federal government has this authority." NIRS argues that NRC logic is simply flaw~d~ The denial of a permit by a State for what;ever reason is not the same as the authority to grant a siting permit or construction license. NRC can welcom~ the consensus process for siting rather than resorting to federal pre-emptive\ authority to,advance utility interests.

NIRS argues that states should and do have tb;e right to deny site permits.

• State. governments are asked to assume many responsibilities with regard to nuclear power plants ranging from "low-level" radioactive waste management to emergency planning.

States therefore have the right to evaluate their resources and balance them with utility interests. NIRS argue~ that States have the right to exercise a more significant role in determining energy resource management in nonconventional fuel sources and energy efficiency and conservation programs for meeting energy needs.

5 0 ~1-t H O0 ('51 Ffc ~7 g-o-;.) NUCLEAR MANAGEMENT AND RESOURCES COUNCIL 1776 Eye Street. NW

• Suite 300
• Washington, DC 200J6-3706 (202) 872-1280 WIiiiam H. Rosin Vice President & Director Technical DMsion March 24, 1993 Mr. Samuel J. Chilk, Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 Attention: Docketing and Service Branch

SUBJECT:

Proposed Rulemaking 10 CFR Parts 50, 52, and 100, "Reactor Siting Criteria" (57 Federal Register 47802 - October 20, 1992 and 55601 - November 25, 1992)

Dear Mr. Chilk:

The Nuclear Management and Resources Council (NUMARC)1, on behalf of the nuclear power industry, has reviewed the proposed rule, "10 CFR Parts 50, 52, and 100, Reactor Siting Criteria," (Federal Register 47802 - October 20, 1992 and 55601 - November 25, 1992) and offers the following comments for consideration. Comments related to the non-seismic part of the proposed rule are presented first, followed by comments related to the seismic part. Non-Seismic The proposed 10 CFR Part 100 revision contained in the Federal Register would codify population density and minimum exclusion area size numeric criteria for siting future nuclear power plants. This NRC action has the potential for significant unintended impacts to both currently licensed and future plants without providing any identifiable improvement to public health and safety. The proposed criteria could inappropriately disqualify a significant number of licensed nuclear power plants sites and otherwise acceptable ~ew sites from availability to host a new nuclear power plant in the future. Furthermore, adoption of the proposed criteria may adversely affect public 1 NUMARC is the organization of the nuclear power industry that coordinates the combined efforts of all utilities licensed by the NRC to construct or operate a nuclear power plant, and of other nuclear industry organizations, in all matters involving generic regulatory policy and on the regulatory aspects of generic operational and technical issues that affect the nuclear power industry. Every utility responsible for constructing and operating a commercial nuclear facility is a member of the NUMARC. In addition, NUMARC's member include major architect-engineering firms and all the major steam supply vendors. Acknowled ed b MAY 1 l .1993 ._: g y card ................."*************n

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Mr. Samuel J. Chilk March 24, 1993 Page 2 perception regarding the acceptable ~ety of existing plant sites during their operating term and during plant license renewal proceedings. An inappropriate message regarding the safety risk associated with advanced light water reactor designs would also be sent when future reactors, which will reflect through their design, construction and operatio~ risk characteristics that are equal to or better than existing plants, are required to meet NRC siting standards that currently licensed nuclear plant sites may not meet. This inappropriate message may also impact the acceptability of siting U.S. advanced reactors in foreign countries, where the availability of remote sites is limited.

• The industry recommends that the radiological dose consequence evaluation factors contained in the current 10 CFR Part 100 be retained as the key determinants of -

site suitability. Codifying in regulation the guidance contained in Regulatory Guide 4.7 (RG-4.7), numeric criteria for minimum exclusion area distance and population density, is inappropriate. This guidance has no demonstrated technical basis and does not reflect the accumulated experience of operating reactors and studies performed- by the NRC and the industry since 1975, the year this guidance was adopted in RG-4.7. In a July 8, 1992, letter to the Chairman, NUMARC urged the Commission to ensure that the proposed rulemaking contain siting criteria with sound technical bases and we strongly urge the Commission not to proceed without such bases. This letter and the enclosures provide our recommendations, answers to the eight questions posed in Part A of Section XI of the October 20, 1992, Federal Register, and the basis for our position (see Enclosures 1, 2, 3, and 4). We believe that criteria contained in the current Part 100, successfully used to safely site all licensed nuclear power reactors in the United States, have the prerequisite technical basis, provide for adequate protection of public health and safety, and are appropriate for the determination of exclusion area distance, low population zone, and population center distance of future nuclear power plant sites. A full discussion of our position is delineated in Enclosure 1, the key elements of which are:

• Improved understanding of postulated severe accident phenomena, probability, and consequences demonstrates that the existing criteria provide significantly less radiological dose risk to the public than previously accepted prior to the 1979 issuance of NUREG-0625. -
• Current licensed plants meet the Quantitative Health Objectives (QHO) of the NRC Safety Goal Policy by a wide margin. Therefore, adequate defense in depth is maintained using the current Part 100 criteria for next generation plants.

Mr. SamuelJ. Chilk March 24, 1993 Page 3

• The next generation nuclear power plants will have considered postulated severe accidents ( class 9) within their designs in accordance with NRCs published Policy Statement on Severe Reactor Accidents Regarding Future Desigru and Existing Plants (50 Fed. Reg. 32138 - August 8, 1985).

Therefore, future reactors will reflect through their designs, construction and operation, risk characteristics that are equal to or better than existing plants.

• 10 CFR Part 52 provides for decoupling of the site suitability determination from the design. Design certification is a completely separate proceeding from siting, therefore the design is approved by the NRC without site specific considerations. Siting accomplished through an early site permit or a combined license will consider the suitability of a site to host a standard certified design on that site independent of the design approval. These processes also provide for the consideration of demographic criteria associated with population density and distnbution in a way that is most meaningful for assuring public safety utilizing defense in depth strategies, namely, emergency planning requirements.
• The proposed criteria, as demonstrated in the enclosures, do not influence conformance with the NRC Safety Goal Policy and provide no significant benefit to protection of public health and safety.

For these reasons, we request NRC reconsider the proposed non-seismic rule, since a sound technical basis for changing the current criteria has not been provided. We also request that the existing siting requirements, with their demonstrated record of safe siting determinations, be retained. Seismic The NRC, by locating much of the prescriptive guidance for the seismic considerations in the regulatory guides has moved in the proper direction with respect to stabilizing the regulation. The industry, however, has several concerns with the language and intent of the proposed rule and associated guidance documents. The industry response to the proposed rule and guidance documents is provided in the form of line in/line out of the original text contained in the proposed rule and its associated guidance

Mr. SamuelJ. Chilk March 24, 1993 Page 4 documents. These modified documents and answers to the five questions posed in Part B of Section XI of the October 20, 1992, Federal Register are included with this letter as Enclosures 5 through 12. Industry's major concerns, which are reflected in the modifications shown in the aforementioned enclosures, ai::e discussed below~ One of the major concerns with the proposed seismic siting rule is related to the requirement that "both deterministic and probabilistic evaluations must be conducted to determine site suitability and seismic design requirements for the site." For the past year, the industry has maintained that a probabilistic approach alone is better suited to satisfying the intent of the rule. NUMARC expressed this concern via letters from Mr. W.H. Rasin to Mr. E.S. Beckjord dated May 8, 1992 and from Mr. Byron Lee, Jr to Chairman Ivan Selin dated June 22, 1992. We also expressed this concern during public meetings with the NRC staff on June 17, 1992 and September 11, 1992~ NUMARC strongly recommends that the proposed dual evaluation requirement be removed from the rule. It is our belief that requiring both types of evaluations to be conducted in parallel is fundamentally flawed Since there is no clear technical procedure for reconciling any differences between the two evaluations, this requirement would most likely destabilize the siting process. The potential for regulatoty instability is contrary to NRC staff's Objective Number 2 identified in the preamble to the proposed rule change (Federal Register 47803 - October 20, 1992). The industry recognizes that in the June 24, 1992, NRC Commission Briefing on the Proposed Part 100 Rule Change the NRC staff stated that even though the language in the proposed rule requires a parallel deterministic and probabilistic analysis, the intent of the rule is not to require a dual process, that an integrated approach similar to the alternative integrated seismic siting decision process. developed by the industry would satisfy the intent of the rule, and that the NRC staff would work with the industry to incorporate appropriate wording in the rule to eliminate the dual evaluation requirement. The industry f'irni1y believes that the probabilistic methods are quite mature and well understood for performing site investigations. Also, probabilistic evaluations account for seismic source interpretation uncertainties whereas deterministic evaluations are simply one interpretation of a range of alternative interpretations. Therefore, we strongly recommend that the requirement for performing a deterministic analysis be removed from the rule. It is our understanding that a key reason for including a deterministic evaluation is the concern that new paleoseismic data may not be accounted for in a probabilistic evaluation since there is. currently no estaplished procedure for updating the existing, accepted seismic hazard studies. In order to address this concern, the industry developed

Mr.Samuell. Chllk March 24, 1993 Page 5 an alternative seismic siting decision process that accounts for new paleoseismic information through a systematic geological, seismological, and geophysical investigation. This process, which is based on a probabilistic approach, has been verified for robustness via an application to a hypothetical site in the Wabash Valley. Details of the industry's alternative process have. been presented through five public meetings between NUMARC and the NRC staff during the past year. We believe that the methodology is robust and technically sound and that it should be given serious consideration as an alternative to the probabilistic evaluation outlined in the proposed rule and its associated guidance documents. The industry believes that making the Operating Basis Earthquake (QBE) applicable only to plant shutdown decisions is an appropriate improvement We, however, have a concern that in many regions in the Eastern US, where the site Safe Shutdown Earthquake (SSE) is less than the standard 0.3g ALWR plant design SSE, assigning the QBE a value of 1/3 of the site SSE may be unn~cessarily restrictive and may require an unnecessary seismic analysis if the applicant chooses to demonstrate acceptability of an QBE value that is greater than 1/3 of the site SSE. We recommend that for sites where the site SSE is less than the standard design SSE, the applicant be able to choose a value of the QBE that is greater than 1/3 of the site SSE, without performing the seismic analysis outlined in the proposed rule (Appendix S to 10 CFR Part 50). The basis for establishing a minimum QBE value of 1/3 of the SSE did not evolve from purely technical arguments, but rather from experience. We believe that our recommendation is sound, especially in light of the fact that in Germany, as stated by R. M. Kenneally of the NRC RES staff ("Revision of Seismic Design Regulations," Proceedings of the Fourth Symposium on Current Issues Related to Nuclear Power Plant Structures, Equipment and Piping, Orlando, FL, December 1992), the latest regulation requires an inspection level earthquake (for shutdown) of 0.4 of the site SSE. Another major concern is related to the determination of QBE exceedance for plants when the seismic instrumentation is inoperable (see Enclosure 10, DG-1017, Appendix A). We believe that basing the QBE exceedance determination on a Modified Mercalli Intensity (MMI) could result in unnecessary plant shutdowns. Instead, we recommend that the earthquake intensity determination be based on EPRI Damage Intensity (EPRI NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake"). The intensity determination outlined in the EPRI report is directly applicable to nuclear plant structures whereas the MMI is more applicable to conventional and residential buildings.

Mr. Samuel]. Chilk March 24, 1993 , Page 6 Snmmm In conclusion, the industry observes that as a whole the proposed rule change package is contradictory in several respects. NRC staff is proposing less prescriptive language in the regula~ons and placing the more specific seismic criteria and guidance in several regulatory guides. In a reversal of this approach, non-seismic criteria and guidance previously contained in RG-4.7 is being proposed for codification. The seismic portion of the proposed rule permits applicants to engineer and to safely accommodate . unfavorable site features whereas the non-seismic portion (demographics) preclude consideration of plant design to offset the prescnbed exclusion area and population. Also, the seismic relaied requirements employ a technical basis in managing risk to the public, whereas the proposed non-seismic requirements lack a firm technical basis and have little or no relation to risk. Demographic issues are stressed at the expense of safety analyses which would more appropriately address risk to the public. These concerns are of critical importance to the industry. We would welcome further, public dialogue with the staff to more fully explain our comments or answer any detailed questions. Sincerely, f//4-,~-(-, William H. Rasin DJM/JPR/ Enclosures c: Mr. James M. Taylor, EDO, NRC (w/o enclosures) Mr. Eric S. Beckjord, Director, RES (w/o enclosures) Dr. Thomas E. Murley, Director, NRR (w/o enclosures) Dr. Andrew J. Murphy, RES (w/ enclosures and diskette of enclosures 5 thru 12) Mr. Leonard Soffer, RES (w/ enclosures)

Enclosure 1 NUMARC COMMENTS (NONSEISMIC) NRC Notice of Proposed Rulemaklng: Reactor ~iting Criteria (57 Fed. Reg. 47802, October 20, 1992) Based on our evaluation of the proposed rulemaking in the area of radiological consequence evaluation factors for site suitability determinations, the nuclear power industry recommends that radiological dose consequence evaluation factors contained in the current 10 CFR Part 100 be retained as the key determinants for site suitability. The radiological consequence evaluation factors are as follows:

• Determination of exclusinn area distance shall be~ ".An individual located at any point on the boundary of the exclusion area for two hours immediately following the oruet of the postulated fission product release would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. "

                                                                                        \

Determination of the low population zone shall be; ".An individual located at any point on the outer radius of a low population zone who i.s exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation _dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure. " The nuclear industry recommends that the current 10 CFR Part 100 criterion

• regarding the placement of nuclear power plants away from densely populated centers (population center distance) be retained in regulation as the key determinant for specifying demographic criteria for site suitability. This demographic criterion is as follows:

The population center distance shall be; "the nearest deruely populated center containing more than about 25,000 residents must be located no closer than one and one-third times the outer radius of the low population zone." The industry basis for these recommendations are as follows. Accumulated Operatini Histozy and Advancements in Desiw Analysis The NRC siting regulations, if changed, must properly consider and reflect the approximately 30 years of accumulated commercial nuclear power experience. The Draft Regulatory Analysis properly states that significant changes and advancements in reactor technology have occurred since 1962 (the issue date of the current Part 100), such as improved ability to clean up fission products and to mitigate the consequences of postulated accidents. Much experience bas also been gained in siting nuclear plants since then; in fact, all of the currently licensed 75 sites have been* safely sited using the current 10 CFR Part 100 approach. Extensive insight has been gained in the understanding of postulated severe accident phenomena, probability and' consequences, and the offsite radiological risk associated with postulated severe accidents (e.g., WASH-1400, NUREG-1150, NUREG-1465, the American Physical Society study on radionuclide release from severe accidents at nuclear power plants, the American Nuclear Society (ANS) study on Source Term, industry-sponsored plant-specific Probabilistic Risk Assessments (PRAs) and Industry Degraded_Core Rulemaking Study (IDCOR) report). These studies indicate that the risk to the public from postulated severe accidents is much less than previously believed. The probability of such postulated severe accidents at a typical operating commercial nuclear power plant is -extremely low, and the studies concluded that the likelihood is remote that, early fatality would result if a severe accident did occur. Specifically, in 1982 the American Nuclear Society chartered a ~pecial Committee on Source Terms to examine the state of knowledge relative to the source teITIL The 15-member Special Committee represented a broad range of knowledge, with representatives from the industry, the government, the national laboratories, academia and the international community. As reflected in its report issued in 19851, the Committee concluded that, "the state of knowledge and the analytical methods and assumptions on which current calculations of the source term are based have progressed far beyond those on wbich'WASH-1400 (The Reactor Safety Study, 1975) was based." The Committee also found that reductions in the source term from estimates reported in WASH-1400 could range from more than a factor of ten to several factors of ten for the critical fission products in most of the accident scenarios recently considered. They also concluded that, "early containment breach from rapid pressure surges or explosions was found to be sufficiently improbable to warrant its neglect as a significant contributor to source terms."* In 1984 the nuclear industry concluded a three and a half year industry study of degraded core 'rulemaking (IDCOR)2. The study was sponsored by sixty-three nuclear 1 Report of the Special Committee on Source Terms, American Nuclear Society, September 1984. 2 Nuclear Power Plant Response to Severe Accidents, IDCOR Technical Summary Report, November 1981. utilities, architect-engineers and manufacturers of light water reactors in the U.S., as well as groups and agencies in Japan, Sweden and Finland. The full study consists of forty-eight individual documents. The study examined in detail how nuclear plants would behave during a core-damage or fuel-melt accident. ~twas based on detailed examination of four nuclear plants. In the 230-page Technical Summary Report, IDCOR drew the following conclusions:

• The probabilities of severe nuclear accidents occurring are extremely low;
• The fission product source terms - quantities and types of radioactive material released in the event of severe accidents - are likely to be much less than had been calculated in previous studies;
• The risks and consequences to the public of severe nuclear accidents are significantly below those predicted by previous studies; and
• Major design or operational changes in reactors are not warranted.

In 1983, the American Physical Society forined a study group on radionuclide releases from postulated severe accidents at nuclear power plants. The final report was issued by the American Physical Society in February 1985 and the NRC Commissioners were briefed on the results of the study on February 21, 1985. The study group concluded that considerable progress had been made since the publication of WASH-1400 in developing both a scientific basis and calculational ability of predicting the source term. It was concluded that, "In a number of cases, new calculations indicate that the quantity of radionuclides that would reach the environment is significantly lower than that calculated in the Reactor Safety Study." Regarding containment failure, it was noted that, "It is now generally believed that large scale failures of reactor containments will not occur until their yield stresses are exceeded.... " It was also concluded that large early containment failures predicted for certain sequences do not occur if the containment is as strong as calculated because, "accident-induced pressures within the containment are not expected to exceed yield stresses until many hours after the reactor pressure vessel failure" and, "steam explosions large enough to challenge the containment directly are now considered unlikely." Specific studies were also completed by the NRC. In a paper given in May 19843, it was noted that early siting requirements were developed when the commercial power plants themselves were being developed and that in the mid-1970's, WASH-1400 3 Impact of New Knowledge on Nuclear Power Plant Siting and Emergency Procedures, Robert M. Bemero, Director, Accident Source Term Program Office, USNRC; AAAS Annual Meeting, New York, May 24-29, 1984. introduced a new perspective. The results of recent studies were presented where, for the siting studies previously mentioned in the paper, a simplified set of accident releases, Siting Source Terms (SST), were developed. The three releases discussed were an unmitigated core melt in which the containment fails early (SST-1), a partially mitigated core melt in which the containment fails but some containment systems function (SST-2) and a well mitigated core melt in which the containment generally holds (SST-3). As a result of these studies, the curves illustrated that, except for the very worst circumstances, the focus of attention during an emergency should be on the first 2 or 3 miles around the plant . At the ANS/ENS International Topical Meeting on Probabilistic Safety Methods and Applications held in 1985, the NRC staff presented a paper4 on a graded approach strategy in emergency planning. The study presented in this paper also made use of the severe accidJnt release (ie., SST). It was noted that more recent work provided additional insights that for most releases, time from initiation of the accident to start of release is about 2 hours or more; the present "warning" time is from one-half hour to one day. The analyses demonstrated that, "Protective Actions (both evacuation and sheltering) may be necessary beyond two miles, but appear to have a somewhat lesser urgency." For the most severe releases, "evacu~tion out to about 5 miles (necessary only in the downwind sectors), with sheltering elsewhere and relocation within 4 hours of ground exposure," would generally preclude early fatalities. The industry has completed studies with similar results5. These studies demonstrate that individual risk decreases significantly beyond two to three miles from the plant The key finding of the studies is that the postulated amount and types of radionuclides that would be dispersed outside a plant's boundaries in a severe accident are likely to be much less_ than had been assumed based on earlier studies (i.e., Technical Information Document (TID) 14844, dated March 23, 1962). The early studies* had the effect of establishing the license design basis source term for currently operating plants. The TID source term has as its basis large quantities of fission products ( as fraction of total reactor core inventory), which were unrealistically assumed to be released. Two fission product boundaries (fuel cladding and coolant pressure boundary) are assumed to have failed simultaneously and instantly, allowing the fission products to 4 An Examination of a Graded Response Strategy in Emergency Planning and Preparedness, Soffer, L, Martin, J.A and Grill, R.P., Reactor Risk Branch, Office of Nuclear Regulatory Research, USNRC, ANS/ENS International Topical Meeting of Probabilistic Safety Methods and Applications, 1985. 5 Graded Response: The Preferred Evacuation Strate~ for Nuclear Power Plants, NUMARC/NESP-005, February 1989. escape to the containment without credit taken for decay of radionuclides or natural removal mechanisms. The containment is assumed to leak to the environment at its maximum design allowable rate. The later stlidies more accurately model the quantity and composition of the fission product inventory within a reactor core, make use in the models of time progression sequences for fission product boundary failures, and account for radioactive decay and natural removal mechanisms (e.g., chemical plate out/deposition of fission products on surfaces). When engineered safeguard features are introduced into the model, there is no credible failure scenario that can result in an early large release to the environment. The current regulations require an applicant to assume that the large release of fission products to the atmosphere does occur, no matter how improbable, and based on that assumption provide for the public's safety with site features. Therefore, the bases used to achieve safe siting for current operating plants have demonstrated to be very conservative and consistent with defense-in-depth principles. The next generation of advanced reactor designs will have comprehensively considered 30 years of accumulated nuclear plant operational experience, including the consideration of severe accident prevention and mitigation within their designs consistent with NRCs published Policy Statement on Severe Reactor Accidents Regarding Future De.signs and Existing Plants (50 Fed. Reg. 32138 - August 8, 1985), and therefore their design, construction and operation, will result in risks as low or lower than those for existing plants. The proposed rule would codify very conservative numeric criteria for minimnm exclusion area size and population density and would establish these criteria as key indicators of site suitability without regard for offsite radiological risk. The proposed rule bas no demonstrated technical basis'. It neither reflects the extensive siting experience to date which demonstrates that the current requirements and use of improving models have provided very conservative criteria for site suitability nor the studies performed by the NRC and the industry since 1975, the year this guidance was adopted in RG-4.7. Coilifying the proposed numeric criteria would yield negligible, if any, improvement in protection of public health and safety. The NRC Established Safety Goals In 1990, the Commission clarified its policy by establishing what is considered acceptable public risk to be used as bases for future rulemaking initiatives (SRM to SECY-89-102, Implementation of the Safety Goals). In support of these industry comments on the NRC proposed change to Part 100, a study was performed to determine the effect that the NRC proposed rule's population density and exclusion area size criteria might have meeting the Quantitative Health Objectives (QHO) associated with the NRC Safety Goal Policy. In addition, the individual and societal risks from postulated accidents (based on NUREG-1150 methods) were determined and compared to the quantitative prompt and latent health objectives of the NRC Safety Goal Policy. The industry study report is provided as Enclosure 4. This evaluation concludes that a nuclear power plant of current design located in an area of controlled population distribution (500 people per square mile) and having a conservative exclusion area distance (033 miles, the actual size for Surry), results in a calculated societal radiation dose risk from postulated severe accidents of approximately 13 person-rem per reactor-year. The calculations of individual radiological risks from postulated accidents demonstrate that the quantitative prompt and latent health objectives of the NRC Safety Goal Policy are met by a wide margin. The prompt risk is shown, to be at its maximum, a small fraction of the QHO, to drop rapidly with distance and is insignificant beyond five miles. The individual latent risk meets the QHO by a wider margin, with a* maximum individual risk of about 1% of the QHO. As with the prompt risk, the individual latent risk initially declines rapidly with distance, with the risk reduced to about 10% of its maximum value at ten miles out. The evaluation also indicates that alternative population configurations with larger overall populations may result in lower societal radiation dose risk, yet fail to meet the NRC proposed criterion. This may result in the e1irnination of otherwise acceptable (possibly the more favorable) sites from further consideration, while less favorable sites survive the site selection screening process. The next generation of nuclear plants will have comprehensively considered in their designs Severe accident prevention and mitigation, as well as accumulated knowledge regarding postulated severe accident phenomena, probability, and consequences. Therefore, these advanced nuclear plants will have risk characteristics that are equal to or better than the existing plants. This evaluation demonstrates that the proposed population density and exclusion area size criteria is unnecessarily restrictive. Providin~ a Stable Regulatory Basis The stated purpose for the proposed 10 CFR Part 50 and 100 (nonseismic) rulernaldng "is to provide a stable regulatory basis for the siting of nuclear power plants by decoupling decisions of site suitability from those affecting plant design." The proposed rule suggests that this objective will be accomplished by having the source terms and dose calculations that are currently associated with 10 CFR Part 100 to establish the site exclusion area size, the low population zone size, and the nearest population center distance, made a part of 10 CFR Part 50 (§ 5034), with some changes from that currently contained in 10 CFR Part 100, and by providing numeric criteria in 10 CFR Part 100 for site minimum exclusion area distance (0.4 miles) and population density criteria (the codification in regulation of existing Regulatory Guide 4.7 guidance). Also, 10 CFR Part 52 (§ 52.17(a)(l), Content of Application) would be modified to read, "The application must also contain a description and safety assessment of the site on which the facility is to be located with appropriate attention to features affecting facility design. The assessment must contain an analysis and evaluation of the major structures, systems, and components of the facility that bear significantly on the acceptability of the site under the radiological consequence evaluation factors identified in§ 50.34(a)(l) of this chapter. Site characteristics must comply with Part 100 of this chapter." This approach, if codified, would result in requiring an early site permit (ESP), construction permit (CP), or combined license (COL) siting applicant to determine, as part of the application, the radiological consequences currently required by 10 CFR Part 100, now relocated to § 50.34. Furthermore, no ~tter how excessively conservative the demonstrated result of the evaluation, the proposed regulation (10 CFR Part 100) would impose the surrogate criteria (minimum exclusion area size and population density). The proposed rule requires both the offsite radiological consequence evaluation and meeting the surrogate criteria codified in regulation, previously provided as guidance in Regulatory Guide 4.7. *

• Early Site Permits, Subpart A, 10 CFR Part 52 require an application to contain site suitability determinations, and assessment of impediments to establishing an emergency plan. Since the actual plant design may not be known at ESP stage, postulated plant design parameters that are representative of the actual design(s) may be used to determine site suitability to host a future nuclear power plant. At COL

(§ 52.79(a)(l)) the applicant is required to demonstrate that "the design of the facility falls within the parameters specified in the early site permit." Thus the ESP already provides decoupling of siting from design, and also provides criteria for population density and distribution as part of emergency planning impediments consideration. For COL sitings, where determination of a site's acceptability to host a standard certified design is made, decoupling of siting from design has been achieve~; the design has been approved by the NRC in separate proceedings. The facility emergency plan, with appropriate demographic considerations, will also be part of the COL proceedings. Therefore, the current 10 CFR Part 52 decouples siting from design and incorporates demographic criteria without the need for additional rulemaking. Early Site Permit Renewal The proposed 10 CFR Part 100 requires an ESP renewal applicant to demonstrate that the population density is consistent with the numeric criteria, i.e., 500 persons per square mile averaged over any radial distance out to 30 miles, and the projected population density 40 years after the ESP renewal of 1000 persons per square mile. Since population density limits are not the key determinants of offsite radiological dose risk, this reconsideration of population density (and potential reopening of the entire alternative site selection process) at ESP renewal is inappropriate. Before a plant with an ESP or CP can begin operation the NRC must grant an operating license (OL) or a COL The proceedings to obtain an OL or COL require consider~tion of any significant new information not previously considered in the ESP or CP, including changes regarding offsite radiological dose consequence evaluation factors. A discussion of the nuclear industry's recommendation regarding the proposed changes to Appendix Q of 10 CFR Part 52 is provided in Enclosure 3. Updatin& Existin2 Guidance

      , Regulatory Guide 4.7, and other NRC documents as needed, should be revised to provide guidance consistent with the latest accepted knowledge regarding postulated severe accident consequences. The guidance provided for population density should remain as guidance in RG-4.7, be limited to involving alternative sites demonstration and be changed to guide ail ESP, CP, or COL siting applicant to consider effects only within an area having a 10 mile radius or less from the center line of the reactor, for the

• following reasons:

( 1) The radiological dose risk to the public and the potential health effects to an individual from a postulated severe accident has been demonstrated by numerous independent responsible studies to be sufficiently low beyond 10 miles that it can be considered negligible; (2) The Commission has determined in its Safety Goal Policy that the quantitative health objectives should be based on an area with a one mile radius from the reactor for prompt effects, and for latent health effects on an area with a ten miles radius. The radiological dose risks for prompt and latent health effects beyond these distances have been determined by the NRC to be sufficiently low and, therefore, can be considered negligible; (3) Other NRC regulations presently require the applicant, state and local agencies to prepare and exercise emergency planning activities within the plume exposure pathway ( emergency planning zone). The emergency planning zone (EPZ) is specified by regulation 10 CFR § 50.47(b)(2) to be an area about 10 miles in radius from the reactor. H NRC requires or suggests population density consideration beyond the EPZ, a confusing message will be sent to state and local agencies and to the public that there are population groups beyond the EPZ at existing sites or at future sites that need to be protected in an emergency, and that no ~ctions have been considered for their protection; (4) The nuclear industry is not aware of studies that conclude the existence of health effects that cannot be acceptably mitigated resulting from long term land contamination from a postulated severe accident. The NRCs Safety Goal Policy and regulations are silent on this issue. The potential economic and environmental impacts from the loss of Ian~ use due t°" land contamination are considered in NRC determinations, required by the National Environmental Policy Act (NEPA), that consider potential environmental impacts in relation to potential plant design features for the prevention and mitigation of severe accidents. It is the nuclear industry's recommendation that loss of lanq use due _to contamination from a postulated severe accident remain an environmental consideration that will be appropriately

• evaluated and considered for each standard plant design during the design certification proceedings. Therefore, land contamination should not be used as a determinant for population density guidance associated with site suitability; and (5) The next generation of advanced nuclear plants will have comprehensively considered in their designs severe accident prevention and mitigation in accordance with NRCs published Policy Statement on Severe Accidents Regarding Future Designs and Existing Plants (50 Fed. Reg. 32138 -

August, 1985). Therefore, the next generation of nuclear plant designs will demonstrate during their design certification proceedings that radiological dose risk and postulated severe accident consequences offsite will be equal or better than currently operating plants. The Federal. Register states the rationale for why meteorological data are no longer needed to be calculated for the purpose of determining site suitability. It states, "analysis reported in NUREG/CR-2239, Technical Guidance for Siting Criteria Development, dated December 1982, for example, show that calculated average individual consequences for an identical postulated release of radioactivity to the environment using data from weather stations throughout the United States yielded results that varied only by about a factor of two. Based upon these considerations, the Commission has determined that the*

• average meteorological characteristics between one site and another are sufficiently similar that characterization of individual site meteorology is not significant discriminator in determining site suitability wben compared to the uncertainties in other areas of the determination of risk to tbe health and safety of the public." The proposed§ 5034 (a)(l) and Regulatory Guide 4.7 would still require the collection of one year's site specific meteorological data and calculations to determine the site atmospheric dispersion values. Regulatory Guide 4.7 and other NRC documents should reflect the benefits afforded by the 10 CFR Part 52 proces.s, the standardization of future nuclear plant designs, and the results of these NRC studies on meteorological conditions.

The standard designs will be certified by the NRC in accordance with Subpart B of 10 CFR Part 52 and thereby have demonstrated during those procee~gs that conservative meteorological parameters were used as basis for the design. The

  , atmospheric dispersion (Chi/Q) design values, as stated in the Utility Requirement Document, are "the Chi/Q values to be used for the licensing offsite dose evaluation and were determined using meteorological data representative on an 80th - 90th percentile U.S. site. The Chi/Q values were calculated following guidance in Regulatory Guide 1.145 considering ground level release, building wake (building area of 33,000 square feet), and lateral plume meander under stable atmospheric conditions." These parameters were chosen so that advanced nuclear plants can be hosted by sites in most

• areas of the United States without jeopardizing the standardization of the designs. The
• radiological dose consequence evaluation factors will have been satisfied by each standard reactor design during its design certification proceedings based on conservative meteorological data.

The current and the proposed regulation require the applicant to determine atmospheric dispersion rates at the site. These values are derived from accumulated meteorological data collected for at least one year at the site (experience bas shown that often substantially more than one year is required to collect this data in compliance with the specified reliability). This data typically can only be collected by the use of a meteo~ological tower erected on the site. In the past the collection of meteorological data bas frequently been the critical path activity in characterizing the site and preparing an application for nuclear power plant site approval.

• An application that utilizes actual standard plant design parameters or an envelope based on standard certified designs will have demonstrated compliance with the radiological dose consequence evaluation factors consistent with the current 10 CFR Part 100 requirements, and therefore site suitability determination is possible.

The meteorological investigation that remains necessary for site suitability determination can be limited to characterization of severe weather phenomena that has the potential for existing at the site, and verification that the site atmospheric dispersion (Chi/Q) is boµnded by the conservative Cbi/Q used for the design. These needs can be accomplished by investigation of "readily available" information such as provided in scientific literature, reports of government and private agencies, consultation with experts, and brief field investigations. The determination of the refined Chi/Q for the site is required for reasons other than the site suitability decision, e.g., determining the appropriate set points for radiation monitoring devices that provide alarms and signals for automatic termination of plant radioactive effluent discharges. If needed, the one year accumulation of meteorological data could be collected during the construction of the plant, thus removing it from unnecessarily delaying plant siting determinations.

Conclusions The proposed 10 CPR Part 100 revision contained in the Federal Register would codify population density and minimum exclusion area distance numeric criteria for siting future nuclear power plants. Discouraging siting of nuclear power plants near populated areas is a policy choice based on societal judgments, it should not be b&5ed on numeric criteria associated with considerations of alternative sites. This NRC action has no demonstrated technical basis and has the potential for significant unintended impacts to both current and future plants without providing any identifiable improvement to public health and safety. The proposed criteria could inappropriately disqualify a significant number of licensed nuclear power plant sites and otherwise acceptable new sites from availability to host a new nuclear power plant in the future. Furthermore, public perception regarding the acceptable safety of existing plant sites may be adversely affected during their operating term and during plant license renewal proceedings. An inappropriate message regarding the safety risk associated with advanced light water reactor designs would also be sent when future reactors, which will reflect through their design, construction and operation, risk characteristics that are equal to or better than existing plants, are required to meet NRC siting standards that currently licensed nuclear plant sites may not meet. This inappropriate message may also impact the acceptability of siting U.S. advanced reactors in foreign countries, where the availability of remote sites is limited. Enclosure 2 NUMARC ANSWERS TO NRC QUESTIONS NRC Notice of Proposed Rulemaking: Reactor Siting Crl~erla (57 Fed. Reg. 47802, Section XI-A - October 20, 1992) Question Number 1 Should the Commission grandfather existing reactor sites having an exclusion area distance less than 0.4 miles ( 640 meters) for the possible placement of additional units, if those sites are found suitable from safety consideration? Answer Grandfathering, which is necessary if a new approach to siting is required, would be unnecessary if the existing siting requirements are maintained. Siting requirements for future power reactors should achieve a level of acceptable safety that is consistent with requirements for currently licensed plants. Currently licensed plant sites have demonstrated acceptable safety for their reactor designs. The placement of additional units_ of advanced designs on a site should be determined on the basis that safety is maintained as a result of operating all the licensed units on that site. This same basis should be utilized for determining acceptability of unit placement on a site not occupied by an existing unit. The nuclear power industry believes that radiological dose consequence evaluation factors in the current 10 CFR Part 100 are the key and appropriate determinants for site suitability to host additional reactors on a site and that these determinants should be maintained in the rule. Question Number 2 Should the exclusion area distance be smaller than 0.4 mile (640 meters) for plants having reactor power levels significantly less than 3800 Megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example, 0.25 miles or 400 meters)? Answer The exclusion area distance should be determined based on criteria contained in the current 10 CFR Part 100, since power level is not the sole determinant of risk. The current criteria provide the more appropriate measure for determining exclusion area distance for reactors with power levels- significantly smaller than 3800 Megawatts (thermal) or for other site or plant specific factors which might affect dose. Twenty-five of the seventy-five currently operating sites have an exclusion area distance less than the 0.4 miles proposed. Each site has been determined by the NRC to adequately provide for the protection of public health and safety. The flexibility should be retained for future sites to establish exclusion area size smaller than 0.4 miles when a smaller area provides adequate protection of public health and safety as determined by radiological dose consequence evaluation for that site. 4t The nuclear industry recommends that a suggested minimum exclusion area distance of 0.25 miles (400 meters) be adopted in Regulatory Guide 4.7 in place of the current 0.4 miles (660 meters). Based on MELCOR Accident Consequence Code System (MAACS) calculations for prompt fatality consequences of postulated severe accidents, an exclusion area distance of 0.25 miles (400 meters) has been found to meet the quantitative health objective of the NRC Safety Goal Policy. This analysis also showed that the size of the exclusion area is affected not only by reactor power level, but also by plant design features, and site related factors. Therefore, future nuclear pow~r plants will be guided to a minimum 0.25 mile exclusion area distance, but regulated to the current 10 CFR Part 100 requirements. Question Number 3 The Commission proposes to codify the population density guidelines in Regulatory Guide 4.7 which states that the population density should not exceed 500 people per square mile out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter. Comments are specifically requested on questions 3A, 3B, and 3C given below. Question Number 3A Should numerical values of population density appear in the regulation or should the regulation provide merely general guidance, with numerical values provided in a regulatory guide? Answer Population density numeric limits should not be codified in regulation because such criteria provide essentially no contnbution to the protection of public health and safety regarding offsite radiological dose risk beyond the immediate area adjacent to the power plant. The NRC has determined that there are no measurable h_ealth and safety impacts to the public from normal operation of a nuclear power plant. NUREG-0880 states, 'For all plants licensed to operate, NRC has found that there will be no measurable radiological impact on any member of the public from routine operation of the plant. (

Reference:

NRC staff calculations of radiological impacts on humans contained in Final Environmental Statements for specific nuclear power plants, e.g., NUREG-0779, NUREG-0812, and NUREG-0854)." The remaining consideration for siting a nuclear power plant is the risk regarding offsite radiological dose from postulated fission product releases. Therefore, the appropriate detern,inant.s for site suitability should remain the radiological dose consequence evaluation factors contained in the current 10 CFR Part 100. Regulatory Guide 4.7 and other NRC guidance documents should he revised to provide guidance consistent with the latest accepted knowledge regarding postulated severe accident consequences and reflect the benefits afforded by the 10 CFR Part 52 process, standardization of future advanced nuclear plant designs, and conclusions of studies that have been performed by the NRC and the industry. See recommendations provided in Enclosure 1 and other parts of this enclosure. Question Number 3B Assuming numerical values are to be codified, are the values of 500 persons per square mile at the time of site approval and 1000 persons per square mile 40 years ~ thereafter appropriate? If not, what other numerical values should be codified and what* is the basis for these values? Answer See answer to question 3A In addition, as stated in the* Federal Register, these criteria should not be considered as an upper limit of acceptability. Much higher population density values have been determined as providing no undue risk to public protection and safety. Codification of requirements to forecast population density values forty years into the future and then compare them to an arbitrary numeric criteria (1000 person per square mile) for site suitability detenninant is inappropriate since such requirements serve no useful purpose in determining risk to the public from radiological dose consequences. When such information is requested in guidance as one of many cost/benefit considerations in the selection of a site from among alternative sites, the projections can be based on bounding calculations, known future plans of local communities, historic trends and the applicant's assumptions. If the forecast requirement were to be codified in regulation as a key determinant for NRC decisions, this forecast could be tested for accuracy beyond its intended importance in license proceedings, and may become the subject of litigation. Question Number 3C Should population density criteria be specified out to a distance other than 30 miles (50 km), for example, 20 miles (32 km)? Ha different distance is recommended, what is its basis? Answer See answer to question 3A In addition, the following comments are provided. Regulatory Guide 4.7 should be changed to recognize that population density within a 10 mile radius or less of the plant is appropriate guidance for the siting of the next generation advanced nude~ power plants. The population density criteria being imposed out to a radius of 30 miles from the reactor is overly restrictive without a commensurate improvement to protection of public health and safety. The Report of the Siting Policy Task Force (NUREG-0625 dated August 1979) stated, based on the judgement of expert members, that no population density limits should be specified beyond 20 miles radius from the reactor because the residual risk to the public was very small and therefore could be considered negligible. Many studies have been performed by the NRC and others since 1979 to better understand postulated severe accident phenomena, probability and consequences (e.g., NUREG-1150, NUREG-1465, Industry Degraded Core Rulemaking Study (IDCOR) report, and industry-sponsored plant-specific Probabilistic Risk Assessments (PRA)). These studies overwhelmingly confirmed

• that early assumptions, including the recommendations presented in NUREG-0625 are unnecessarily conservative.

A study based on NUREG 1150 methods, Evaluation of Population Distribution Relative to Meeting the Quantitative Health Objectives of the NRC Safety Goal Policy for Offsite Risk Associated with Nuclear Power Plants," concluded that nuclear power plants of current design located in an area of controlled population distnbution (500 people per square mile) and minimum exclusion area boundary of 033 miles, result in a calculated societal radiation dose risk for reactor accidents of approximately 13 person-rem per reactor year (Enclosure 4). The calculated individual radiological risks from postulated accidents demonstrate that the quantitative prompt and latent health objectives of the NRC Safety Goal Policy are met by a wide margin, and that there are inconsequential additional benefits to protection of public health and safety gained beyond 10 miles from the reactor plant. Therefore, consideration of the population distribution is more appropriate within 10 mile radius. The demographic criterion contained in the current 10 CFR Part 100 should continue to be the key determinant of acceptable population densities. The precedent has been established by: (1) The NRC Safety Goal policy is based on 10 miles; (2) Current regulations require applicants to provide for public safety by establishing emergency plans for the protection of. the public adjacent to nuclear plants, within about 10 miles *- (Emergency Planning Zone); and (3) Severe accident consequence analyses performed by the NRC and the nuclear industry determined that the health risk to the public is low beyond 10 miles, below the established targets based on risks attributable to other causes. Potential long term land contamination from postulated severe accidents should not be used as a determinant for establishing the size of the area for population density standards or guidance without some basis for believing that remote siting requirements would appreciably ameliorate such effects. The industry is unaware of any such technical basis. Furthermore, there should be full opportunity for public comment on the technical approaches and results pertaining to the relationship of siting standards and land contamination if it is part of the rationale for the final enactment of this rulemaking. Question Number 4 Should the Comnilssion approve sites that exceed the proposed population values of 10 CFR 10021 and if so, under what conditions? Answer Population density numeric criteria should not be a determinant for site suitability since such criteria do not necessarily yield the most favorable site regarding public risk, associated with radiological dose from a postulated fission product release. The Commission should approve sites that the applicant has demonstrated to adequately protect public health and safety using the current radiological dose consequence evaluation factors. This is the approach the nuclear power industry recommends to remain as the key determinant for site suitability. Therefore, there currently exist adequate bases for the Commission's decision regarding approval of sites.

Question Number s Should holders of early site permits, construction permits, and operating license permits be required to periodically report changes in potential offsite hazards (for example, every five years within 5 miles)? If so, what regulatory purpose would such reporting requirements serve? Answer A new requirement for periodic reporting of offsite hazards is inappropriate. Such a requirement is redundant to current requirements (10 CFR 50.71(e)) for operating licensees (OL) to report potential offsite hazards impact on the plant, as the impact affects public health and safety, through the licensee's update and report to the NRC of its Final Safety Analysis Report (FSAR). During the term of early site permits (ESP) or construction permits (CP) there is no regulatory purpose for periodically reporting changes in potential offsite hazards. Before a plant with a CP or ESP can begin operation the NRC must grant an OL or a combined license (COL) (10 CFR 52.79(b)). The proeyedings to obtain an OL or COL require consideration of any significant new information not previously considered in the ESP or CP, including changes regarding offsite hazards. Therefore, at the point where there is a regulatory purpose to have ESP or CP holders consider potential offsite hazards and make NRC aware of those with significant impact, there already exists an effective regulatory requirement. An added reporting requirement would be redundant and inappropriate. NRC should not require performing evaluation or reporting of information without value to the NRC in fulfilling its regulatory mission. Evaluation of offsite hazards that neither have significant impact, nor require NRC or licensee actions would be unnecessary and inappropriate. New and redundant reporting requirements would increase the cost of nuclear generation of electricity without a commensurate increase in protection of public health and safety. Furthermore, the NRC through its audits of licensee activities can in the future, as it has in the past, audit licensees' analysis and disposition regarding safety determinations performed. _ Since the NRC already receives the appropriate information regarding offsite hazards to ful6ll its regulatory mission, any new requirement to report this information would not meet the intent underlying the Paperwork Reduction Act of 1980. Question Number 6 What continuing regulatory significance should the safety requirements in 10 CFR Part 100 have after granting the initial operating license or combined operating license under 10 CFR Part 52? 10 CFR Part 100, 11Reactor Site Criteria" should remain a regulation that specifies the safety requirements for the Commission's evaiuation of the suitability of proposed sites for stationary power and test reactors. Therefore, 10 CFR Part 100 should not have any continuing re~latozy significance beyond the issuance of early site permits or the sitm& portion of combined licenses. Power plant construction and operational safety requirements are specified in other parts of the Commission's regulations, e.g., 10 CFR Part 50, and site safety requirements are specified in the terms and conditions (Safety Analysis Report) of the applicable licenses for facilities. Site safety requirements that have continued significance to the operational phase of the facility are therefore already provided for. Therefore, additional continued 10 CFR Part 100 regulatory significance is not required and would be inappropriate. Question Number 7 Are there certain site meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are the conditions that can not be adequately compensated for by design features?

• Answer The nuclear industry is not aware of any meteorological conditions that can .nQ.t be adequately compensated for by design features. Siting of a nuclear power plant, therefore, should not be precluded based on site meteorological conditions.

The Regulatory Guides 1.145, 1.23, 4.7 and other applicable regulatory documents should be revised. They should not require site specific collection of one years worth of meteorological data to determine atmospheric dispersion values since this data is no longer needed for the determination of site suitability. The required meteorological investigation should be to characterize severe weather phenomena that have the potential for existing at the site and to demonstrate that the site's atmospheric dispersion factor (Chi/Q) is bounded by the Chi/0 values used for the design. This meteorological investigation will be based upon "readily available" information such as provided in scientific literature, reports of government and private agencies, consultation with experts, and brief field investigations. In the past, collection of meteorological data for determining Chi/Q values frequently required more than one year, and bas ~een the critical path activity in site characterization. The radiation dose limits to an individual, as required by the current 10 CFR Part 100 radiological dose consequence evaluation factors, will be determined for each reactor plant design in its design certification. The industry's goal as stated in the Utility Requirement Document (URD) for advanced light water reactors is to require the design to demonstrate that 10 CFR Part 100 exposure limits can be met using the envelope of ALWR plant site design parameters for Chi/Q. The URD states, "the Chi/Q values are to be used for the licensing offsite dose evaluation and were determined using meteorological data representative of an 80th - 90th percentile U.S. site. The Chi/Q values were calculated following guidance in Regulatory Guide 1.145 considering ground level release, building wake (building area of 33,800 ft'), and lateral plume meander under stable atmospheric conditions." The ALWR vendors have stated that plant designs will achieve the URD goal. Therefore, certified plant designs will meet the o:ffsite radiological dose consequence factors. An applicant that utilizes actual plant design parameters or an envelope based on standard certified design will have demonstrated compliance with the radiological dose consequence evaluation factors consistent with the current 10 CFR Part 100 requirements, and therefore site suitability determination is possible. The determination of the refined Chi/Q for the site is required for other than site suitability reasons, e.g., determining the appropriate set points for radiation monitoring devices that provide alarms and signals for automatic t~rmination of plant radioactive effluent discharges. If needed, the one year accumulation of meteorological data could be collected during the construction of the plant Question Number 8 In the description of the disposition of the recommendations of the Siting Policy Task Force report (NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to reconsider any recommendation not adopted and, if so, what are the bases for reconsideration? Answer There are no additional recommendations contained in the report of the Siting Policy Task Force (NUREG-0625), dated August 1979~ that should be reconsidered for adoption. NUREG-0625, contains policy recommendations that may no longer be appropriate because the assumptions underlying those recommendations were based on information that predate the large amount of accepted knowledge about postulated severe accident phenomena, probability and consequences gained since 1979. Enclosure 3 NUMARC COMMENTS (APPENDIX Q, 10 CFR PART 52) NRC Notice of Proposed Rulemaking: Reactor Site Criteria (57 Fed Reg. 47802 - October 20, 1992)

• Oarification of Langui\ie in Appendix o Reiardjni Early Site Permits The proposed rule would also modify 10 CFR Part 52 Appendix Q to provide that an application for renewal o~ an early site permit "is subject to a full early site permit review." Currently, Section 52.31, Criteria for Renewal, provides that the criteria for renewal of an early site permit is compliance with the Atomic Energy Act and the Commission's rules and regulations applicable and in effect at the time the original site permit was issued and any new requirements that the Commission determines provide "a substantial increase" in overall protection of the public health and* safety as long as the "direct and indirect costs of implementation of those requirements are justified in view of this increased protection." Thus, the proposed change to Appendix Q is directly in opposition with the explicit provisions of Section 5231.
• Section 52.13 descnbes the relationship of Appendix Q to Subpart A of Part 52.

Specifically, it states that "Appendix Q applies only when NRC staff review of one or more site suitability issues is sought separately from and prior to the submittal of a construction permit." Thus, not only is the proposed revision to Appendix Q completely inconsistent with the specific regulatory requirement of Subpart A, it also appears to have no applicability to an early site permit renewal because an early site permit holder seeking renewal would not be seeking the separate review of NRC staff on a site suitability issue under Appendix Q, but rather would be seeking a renewal in accordance with Sections 52.29 and 52.31. In conclusion, and because no insight is provided in the Statement of Considerations of the proposed rule for the amendment to Appendix Q, we recommend that the proposed change be deleted. Not only is it inconsistent with the explicit regulatory language in Part 52, but its adoption would serve to undermine the regulatory stability and predictability that is the fundamental goal of the Part 52 licensing process.

Enclosure 4 EVALUATION OF POPULATION DISTRIBUTION RELATIVE TO :MEETING THE QUANTITATIVE HEALTH OBJECTIVES OF THE NRC SAFETY GOAL POLICY FOR OFFSITE RISK ASSOCIATED WITH NUCLEAR POWER PLANTS EARLY SITE PERMIT DEMONSTRATION PROGRAM MARCH 1993

Executive Surnmax:y In this study the proposed U.S. Nuclear Regulatocy Commission (NRC) population density and exclusion zone size criteria were evaluated for their potential effectiveness to meet the quantitative public health and safety objectives associated with the NRC Safety Goal Policy. In addition, the individual and societal risks from postulated accidents (based on NUREG-1150 methods) were determined and compared to the quantitative prompt and latent health objectives of the safety goal policy.

• The proposed revisions to 10 CFR 100 regarding population density and exclusion zone size criteria were evaluated using the MELCOR Accident Consequence Code System (MACCS).

Surry Nuclear Plant radiological and meteorological input data were taken from the NUREG-1150 analysis. Surry was selected for the following reasons: it is one of the five plants evaluated by the NRC in NUREG-1150; it is one of two plants for which NUREG-1150 considered analysis of external events; and its source term data are readily available. This analysis demonstrates that a nqclear power plant of current design located in an area of controlled population distribution (500 people per square mile) and minimum exclusion zone size (0.33 miles), as specified by the NRC in the proposed 10 CFR 100 revision, results in a calculated societal radiation dose risk from reactor accidents of approximately 13 person-rem per reactor-year. The calculated individual radiological risks from postulated accidents demonstrate that the prompt and latent Quantitative Health Objectives (QHOs) of the NRC Safety Goal Policy are met by a wide margin. NUREG-1150 used site-related factors including plant design features (source term), meteorology, population density and distribution, and estimates of the effectiveness of emergency plans. Therefore, by using the NUREG-1150 study methodology, which fully considers population density and distribution, it is possible to compare offsite health effects to the QHOs of the NRC Safety Goal Policy. Further, the population density criterion proposed for codification will not provide the flexibility needed to select consistently the most favorable sites. Alternative population configurations with larger overall populations may result in lower societal radiation risk yet fail the NRC proposed criterion. This may result in eliminating acceptable (possibly the more favorable) sites from further consideration while less favorable sites survive the site selection screening process. The proposed criterion could unnecessarily constrain future activities at existing sites (if they were to be evaluated under the proposed rule), sites that have previously been approved but have never had operating reactors, and new sites that would likely be considered for future reactors. For example, in the NRC Staff briefing to the Commission, the staff stated that eight licensed sites do not currently meet the five hundred persons per square mile population density criterion, and twenty-five licensed sites do not meet the 0.4 mile exclusion area distance criterion. All of these sites have been demonstrated to adequately protect the public health and safety and easily meet the NRC Safety Goal Policy. If codified, then the restrictive numeric criteria would establish a de facto safety goal in regulation that is more stringent than the NRC Safety Goal Policy and is applicable to future plants only. The next generation of nuclear plants will have comprehensively considered 30 years of accumulated experience including improved understanding of postulated severe accident phenomena, probability. and consequences. Therefore. these plants will reflect in their design, construction, and operation risk characteristics as low or lower than those for existing plants.

The proposed criteria would be significantly more restrictive than what is required to meet the QHOs of the Safety Goal Policy and would therefore be unnessarily restrictive. Since the risk to the public would be so much less than the 0.1 % increase from all other factors extablished by ' the QHOs of the Safety Goal Policy, there is insufiicient health risk basis for establishing such a restrictive criteria.

EVALUATION OF THE PROPOSED NRC POPULATION DENSITY CRITERION ON SEVERE ACCIDENT RISK L Summary 10 CFR Part 100, .Reactor Site Criteria, sets forth the criteria for an exclusion ar~ a low populatj.on zone, and a population center distance around nuclear plants. The U.S. Nuclear Rer.latory Commission (NRC) has established population density guidelines in Regulatory Gwde 4.7 to be used as an indicator as to whether a facility meets the Fedentl requirements. The NRC has issued for public comment the proposed change to 10 CFR Part 100 which, if promulgated, would codify these population density and exclusion zone size guidelines. This study was undertaken to evaluate (1) the proposed population and exclusion zone size criteria for their effectiveness in meeting the public health and safety objectives associated with the prompt and latent quantitative health objectives (QHOs) of the NRC Safety Goal Policy and (2) the individual and societal health and safety risks for several population distributions using the methodology used by the NRC in NUREG-1150, Severe Accident Risks: An Assessment for Five U. S. Nuclear Power Plants. The results of this study indicate that population density and exclusion zone size criteria proposed in 10 CFR Part 100 do not influence conformance to the QHOs of the safety goal policy. Further, the results of this study have shown that a reactor of current design located in an area of controlled population distribution and minimum exclusion zone size as specified by the NRC in the proposed 10 CFR Part 100 revision results in a calculated societal risk of approximately 13 person-rem per reactor-year. The study also demonstrates that such a facility meets the individual QHOs by a wide margin. NUREG-1150 used site-related factors including plant design features Jsource term), meteorology, population density and distribution, and estimates of the effectiveness of emergency plans. Therefore, by using the NUREG-1150 study methodology. which fully considers population density and distribution, it is possible to compare offsite health effects to the QHOs of the NRC Safety Goal Policy. However, because population density guidelines would only limit the number of potentially affected individuals in the vicinity of the plant, it d~ not address the benefits of other influences pertinent to protection of public health and safety. - Thus, the population density criterion proposed for codification will not provide the flexibility needed to select consistently the most favorable sites. Alternative population configurations with larger overall populations may result in lower societal radiation risk yet fail the NRC proposed criterion. This may result in eliminating acceptable (possibly the more favorable) sites from further consideration while less favorable sites survive the site selec~on screening process. The proposed criterion could unnecessarily constrain future activities at existing sites {if they were to be evaluated under the proposed rule). sites that have previously been approved but have never had operating reactors, and new sites that would likely be considered for future reactors. For example, in the NRC Staff briefing to the Commission, the staff stated that eight licensed sites do not currently meet the five hundred persons per square mile population density criterion, and twenty-five licensed sites do not meet the 0.4 mile exclusion area distance criterion. All of these sites have been demonstrated to adequately protect the public health and safety and easily meet the NRC Safety Goal Policy. If codified, then the restrictive numeric criteria would establish a de facto safety goal in regulation that is more stringent than the NRC Safety Goal Policy and is applicable to future plants only. The next generation of nuclear plants will have comprehensively considered 30 years of accumulated experience including improved understanding of postulated severe accident phenomena., probability, and consequences. Therefore, these plants will reflect in their design, construction, and operation; risk characteristics as low or lower than those for existing plants. 1

The proposed criteria would be significantly more restrictive than what is required to meet the QHOs of the Safety Goal Policy and would therefore be unnessarily restrictive. Since the risk to the public would be so much less than the 0.1 % increase from all other factors extablished by the QHOs of the Safety Goal Policy, there is in,sufficient health risk basis for establishing such a restrictive criteria. IL Objective To evaluate population distributions adjacent to a nuclear power plant and exclusion zone size relative to offsite risks from postulated accidents and to compare these risks to the QHOs of the NRC Safety Goal Policy using the NUREG-1150 methodology. m Backa,:ound Existing NRC policy_ establishes the safety goals for individual risks in the vicinity of nuclear power plants. These goals define what the NRC considers as adequate safety and risk beyond which the risk is considered to be remote and speculative for new rulemaking. These safety goals state that the risk of prompt or latent {cancer) fatalities to an average individual in the vicinity of a nuclear power plant should not exceed 0.1 % of the risk from other causes. For the prompt risk safety goal, the vicinity is defined as within one mile of the reactor, whereas for the latent risk safety goal the vicinity is defined as within a ten mile radius of the reactor. The NRC has !lefined QHOs corresponding to the above goals as equivalent to 5 x 10-7 and 2 x 10-6 fatalities per reactor-year for prompt and latent cancer fatalities, respectively. I 10 CFR Part 100 specifies the following: An exclusion area A low population zone (LPZ) A population center distance criteria The exclusion area is to be of sufficient size such that any individual in this area (but outside the plant protected area) would not receive, upon two hours of exposure following a postulated release, a total radiation dose to the whole body in excess of 25 rem or a total radiation dose to the thyroid in excess of 300 rem from iodine e~posure. The LPZ is defined such that any individual in the LPZ who is exposed to the entire passage of the radioactive cloud resulting from a postulated release would not receive a dosage in excess of the above limits. , The populatiqn center distance criteria states that no population center {defined as a densely populated area with a total population in excess of 25,000 people) shall be closer than 1 1/3 times the distance from the reactor to the outer boundary of the LPZ. The intent of the LPZ is to ensure that the papulation is sufficiently limited and, furthermore, is distributed in such a fashion that mitigating action {such as an evacuation) would likely be successful. The exclusion area was established based on the premise that successful mitigation might not be achievable given excessive radiation. Finally, the population center distance criterion addresses societal risk (albeit indirectly) in that it has been deemed desirable to keep large populations relatively far from the reactor, thereby minimizing the potential for large scale consequences. 1u.s. Nuclear Regualtory Commission, SECY-89-102, lmplementanon oftM Safety Goal Policy, March 30, 1989, and Staff Requirements Memorandum to SECY-89-102, June 15, 1989. 2

The distances and areas specified above cannot be precisely determined due to variability of release patterns, meteorology, and location of the most exposed individual and are specific to each operating nuclear plant site. The NRC established population density guidelines in Regulatory Guide 4.7 issued in November, 1975._ The NRC detennined that an exclusion area radius of 0.4 miles would be acceptable for all reactors when the Regulatory Guide was published. In addition, the NRC determined that an acceptable surrounding population density should not exceed 500 people per square mile averaged over any radial distance out to 30 miles. Furthennore, the projected population for the operating lifetime of the reactqr (40 years) should not exceed 1000 people per square mile. The probable intent of the above population density requirements was to ensure a sufficiently low population density to allow for effective mitigation. measures, while assuring that population centers would not encroach on the reactor.2 . The NRC is now proposing to codify the above guidelines. The stated objective for this change is to, "Relocate the requirements that apply to plant design into 10 CFR part 50 thereby effectively decoupling siting from plant design."3 Another projected benefit is that by establishing quantitative population rules, the siting requirements, become clearly defined. This, in tum, is expected to reduce the time and costs associated with obtaining site approval. IV. Methodology The population density and exclusion zone siz.e criteria in the proposed revisions to 10 CFR Part 100 were evaluated using the MELCOR Accident Consequence Code System (MACCS), developed by Sandia National Laboratories for the NRC. Surry Nuclear Plant radiological and meteorological input data were taken from the NUREG-1150 analysis. Surry was selected for the following reasons: it is one of the five plants evaluated by the NRC in NUREG-1150; it is one of two plants for which NUREG-1150 considered analysis of external events; and its source tenn data are readily available. Additional discussion of the data and assumptions utilized in the MACCS analysis is provided in Appendix A V. Results This section presents the results of the MACCS analysis of the four release scenarios discussed in Appendix A applied to a uniform population density of 500 people per square mile and an exclusion zone size of 0.33 miles. While MACCS can be used to calculate and display a variety of quantitative infonnation, individual risk probabilities (identical to NRC QHO units) and the population dose (assumed to be a measure of cumulative societal risk) were selected for comparison. For the NUREG 1150 analysis, the Surry source tenns were divided into 17 groups and further divided into three subgroups each for a total of 51 subgroups. For purposes of this analysis, these groups were condensed into four categories, and representative source terms were chosen for each category. Appendix A provides further detail on the assumptions used for this analysis. Figures 1 and 2 present comparisons of the quantitative prompt and latent individual risks from this analysis with their respective QHOs. As can be seen in Figure 1, the prompt risk is well 2Toe draft document for Regulatory Guide 4. 7 issued the previous year made no mention of these population density guidelines. From review of available public documents, it was determined that (1) if these guidelines, which were later added to Regulatory Guide 4.7. had been added without allowing for public discussion and (2) the justification for the selection of the 500/1000 population density criterion (not given in Regulatory Guide 4.7) could not be found. 3u. S. Nuclear Regulatory Commission, October 20, 1992 (57 FR 47803). 3

below the QHO, reaching its peak value of 8.4% of the QHO at the boundary of the exclusion area (assumed for this study to be 0.33 miles). The prompt risk is shown to drop rapidly, and the residual radiation risk is insignificant beyond five miles. The individual latent risk, shown in Figure 2, meets the QHO by even a wider margin..with a maximum individual risk of about 1% of the QHO. As with the prompt risk, the individual latent risk initially declines rapidly with distance with the risk reduced to about 10% of its maximum value at ten miles out The risk then continues to decrease with radial distance. Using a weighted average of all four modeled scenarios, Figure 3 depicts a plot of the resulting dose as it varies with radial distance. This dose is comprised of two components: dose from immediate impact (e.g., cloudshine, groundshine, inhalation, and deposition on skin), and dose from long term impact (e.g., consumption of contaminated milk, water, or food, as well as, dose from decontamination related activities). As with the individual risk, the dose declines rapidly with distance and is reduced to less than 10% of its maximum value at ten miles. Figure 4 shows the immediate, long term, and combined population dose (individual dose weighted with population) as a function of distance. The combined population dose is shown to decline only slightly with distance, indicating that the decrease in individual dose is balanced by the increase in population. Note, however, that the immediate population dose decreases with distance, sharply at first, then declining more gradually. In contrast, the long term cumulative population dose increases with distance for approximately the first 15 miles then becomes stable with distance. The total societal risk to a population (i.e., the cumulative risk to all people in the 30-mile radius) using the algorithms in MACCS is determined by integrating the area under the combined population curve in Figure 4. This assumes that latent risk is proportional to cumulative dose regardless of how the dose is distributed to individuals (e.g., one person-rem divided among 1000 people is equivalent to one person-rem to a single person). Using the NUREG-1150 data for the Surry Plant and ~urning a uniform population density of 500 people per square mile and exclusion zone size of 0.33 miles, the total dose is calculated to be about 13 person-rem per reactor-year. This, of course, is the value of an average exposure taken over millions of reactor-years as the population would only be exposed in the extremely unlikely chance of a core damage event This value represents a valid quantitative measure of the societal risk. Figure 5 presents the results of an analysis performed using two assumed population distributions. The first assumes a uniform population density of 500 people per square mile. The second assumes that most of the population is located between the 20 and 30 mile radial distances. This analysis shows a scenario in which the NRC population density criteria is exceeded, yet the total societal risk is significantly lower than a scenario which conforms to the population density criteria. Figures 6 through 8 present the results of an analysis performed assuming a reduced exclusion zone of 0.2 miles from the reactor with and without evacuation. This analysis shows that, assuming evacuation, the individual risk for the 0.2 mile exclusion zone is slightly higher than the individual risk for the 0.33 mile exclusion zone. The individual risk to the nearby population is greater if there is no evacuation, as noted in Figures 6 and 7, and reflected in the population dose profile (Figure 8). The evacuation was ~urned to be carried out to ten miles only (the same ~umption as NUREG-1150); therefore, the risk to the greater than ten mile population is identical for both evacuation assumptions. 4

VI. Conclusions As demonstrated by the results presented above, it can be concluded that a reactor of current design located in an area of controlled population distribution and minimum exclusion zone size, as specified by the NRC in the proposed 10 CFR Part 100 revision, results in a calculated annual societal risk which is well within the QHO of the NRC Safety Goal Policy. Considering that such a facility meets the individual QHOs by a wide margin, the proposed NRC population density and exclusion zone size criteria are unnecessarily restrictive and therefore, should not be promulgated. In addition, codification of these criteria is questionable because new future advanced nuclear plant designs have added features to lower the risk when compared to the current generation of plants. Further, the population density criterion proposed for codification will not provide the flexibility needed to select consistently the most favorable sites. Alternative p<?Pulation configurations with larger overall populations may result in lower societal radiation risk yet fail the NRC proposed criterion. This may result in' eliminating acceptable (possibly the more favorable) sites from further consideration while less favorable sites survive the site selection screening process. The

• proposed criterion could unnecessarily constrain future activities at existing sites (if they were to be evaluated under the proposed rule), sites that have previously been approved but have never had operating reactors, and new sites that would likely be considered for future reactors. For example, in the NRC Staff briefing to the Commission, the staff stated that eight licensed sites do not currently meet the five hundred persons per square mile population density criterion, and twenty-five licensed sites do not meet the 0.4 mile exclusion area distance criterion. All of these sites have been demonstrated to adequately protect the public health and safety and easily meet the NRC Safety Goal Policy.

If codified, then the restrictive numeric criteria would establish a de facto safety goal in regulation that is more stringent than the NRC Safety Goal Policy and is applicable to future plants only. The next generation of nuclear plants will have comprehensively considered 30 years of accumulated experience including improved understanding of postulated severe accident phenomena, probability, and consequences. Therefore, these plants will reflect in their design, construction, and operation; risk characteristics as low or lower than those for existing plants.

• The proposed criteria would be significantly more restrictive than what is req~d to meet the QHOs of the Safety Goal Policy and would therefore be unnessarily restrictive. Since _the risk to the public would be so much less than the 0.1 % increase from all other factors extablished by the QHOs of the Safety Goal Policy, there is insufficient health risk basis for establishing such a restrictive criteria.

5

FIGURE 1 PROMPT INDIVIDUAL RISK 100.00 90.00 80.00 8 C, 70.00 10.00 ,.t (1) 60.00 8.00 50.00 6.00 0 4.00 1: 40.00 ! 2.00 0. 30.00

0. 00 _ ____;:::....:::------**.._--1 1,- 1---<,

0.00 5.00 10.00 20.00 10.00 0.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor {m,lles)

                         --1 a--  Individual Prompt Risk - - - Safety Goal

FIGURE 2 LATENT INDIVIDUAL RISK 100.00 90.00 80.00 l C, 70.00 1.20 ....i 60.00 1.00 I 50.00 0.80 0 0.60 1: 0.40

*..u  40.00                  0.20 Q.

C, 30.00 o.oo L-~========-==* 0.00 10.00 20.00 30.00 20.00 10.00 0.00 ..._.11:::a:=11=111=a----+---lll-------l--------l-------+----------l-------*1 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor (miles) Individual Latent Risk - - - Safety Goal

FIGURE 3 INDI.VIDUAL DOSE PROFILE 7.00 6.00 i.. 5.00

    -*g 4.00 Q
    'ii J   3.00 00
    ~   2.00 1.00 0.00           5.00               10.00                   15.00          20.00       25.00 30.00 Distance from Reactor (mlles)

I****** *** Immediate Dose * * * *&* * *

• Long Term Dose - - - Combined Dose The modeled core damage accident Is an amalgamation of a range of release scenarios.

FIGURE 4 POPULATION DOSE PROFILE 0.60 c o.so

 -~
 .! ~

e *

 'ia *>- 0.40
 =am ..o                                                                                          . .--

a: t; 0.30 if G> e 8* *... 0.20

                                    -  ~: . : *...

ti ll. 0.10 0.00 - + - - - - - - - - + - - - - - - - - + - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - l 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor (miles) I****** *** Immediate Dose **- ** * *

• Long Term Dose - - - Combined Dose Cumulative population Dose Risk (within 30 miles) equivalent to approximately 13 man-rem per reactor year.

RGURE 5 POPULATION DOSE PRORLE FOR VARYING POP. PRORLE 0.70 i I 0.60 I

           =

CD ._

E j 0.50 I

           'ii ~

i U0.40 , a: = l a, o.3o I I I 0 CD rJ E e 0.20 , ,- - - - - .... - ... - - I Q c.. CD a. 0 0.10 _____________ , , , 0.00 + - - - - - - + - - - - - - - - - - - - - - - + - - - - - - t - - - - - - - - - - - - - - - 1 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor (miles)

                   - - - - - Stratified Pop: 50 people/sq mile 0.3 - 1o miles;                Uniform Pop: 500 people/sq ml!e* 0.3 - 30 miles.

250 people/sq mile 1o - 20 miles; 750 people/sq mile 20 - 30 miles Total exposure: Uniform Population: 13 man-rem/reactor-year Stratified Population: 8.5 man-rem/reactor-year

FIGURE 6 PROMPT INDMDUAL RISK 100.00 90.00 80.00 1C, 70.00 40.00

        -t

l 60.00 30.00

        -c 0

50.00 40.00 20.00 10.00 0.00 1--...!l!::1. . ._,.I--______,___ ........ ~ Q. 0.00 5.00 10.00 30.00 20.00 10.00 0.00 ---1 0.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor (mlles)

              --11a---  0.33 mile exclusion w/  --4*-- 0.2 mile exclusion w/             _ .....,...........,._ 0.2 mile exclusion w/o       Safety Goal evac.                                     evac.                                         evac.

FIGURE 7 LATENT INDIVIDUAL RISK 100.00 90.00 80.00 1t, 70.00 t rn 0 60.00 50.00 6.00 4.00 c N

   *u D..

40.00 30.00 2.00

0. 00 L:~1::.-111=41----------*

0.00 10.00 20.00 30.00 20.00 10.00 0.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor (miles)

        --t111--- 0.33 mile exclusion     ----4*-- 0.2 mite exclusion w/   _...,..__ _ 0.2 mile exclusion w/o - - - Safety Goal w/ evac.                         evac.                               evac.

FIGURE 8 POPULATION DOSE PROFILE 1.00 0.90 c 0.80 Cl E

      ~ -;:- 0.70

E :

ca o

      * >- 0.60 a:  u   0.50 i l~   o.4o -

8 C .. 0.30 w t 0.20 0.10 0.00 + - - - - - - - - - - - - - - - + - - - - - - - - + - - - - - - - - + - - - - - - - - - - - - - - 0.00 5.00 10.00 15.00 20.00 25.00 30.00 Distance from Reactor {mlles)

                     -     0.33 mile exclusion w/ evac. --<1--- 0.2 mile exclusion w/ evac.    --,j*-- 0.2 mile exclusion w/o evac.

Total dose with evacuation = 13.3 man-rem/year (for both 0.2 and 0.33 mlle exclusion zones). Total dose without evacuation = 14.5 man-rem/year (0.2 mile exclusion zone}.

APPENDIX A Data and Assumptions for MACCS Analysis _ In the NUREG-1150 analysis of the Surry Nuclear Plant, the thousands of source tenns were partitioned into 17 groups. Each of these groups were subdivided into three subgroups: Subgroup 1: Evacuation starts at least 30 minutes before the release begins'. Subgroup 2: Evacuation starts between 30 minutes before and 1 hour after the release begins. Subgroup 3: Evacuation starts more than 1 hour after the release begins. While this gives a potential for 51 subgroups, some of these subgroups have zero or nearly zero conditional probability. Nevertheless, this leaves 34 non-trivial subgroups for MACCS input. Rather than evaluating all of these groups, they were combined to a more manageable number of events while retaining the integrity of the analysis. Figure A-1 depicts the partitioning of the 17 groups by early and latent health effects as evaluated in the NUREG-1150 study. These groups were then further organized into four categories as shown. NUREG-1150 lists the expected annual frequencies of each of the 17 source tenns, the conditional probability of each of their subgroups, and the predicted consequences of each of their occurrence. Table A-1 presents this data condensed into the four categories used for this study, along with the source terms selected to represent each of these categories. With the exception of source tenn 17-1, each of the selected postulated releases closely matches the expected impact of their respective categories. Source tenn 17-1 was conservatively selected to represent its category as it was the only source ~nn in this group with a non-negligible impact Once the representative release scenarios were selected, it was necessary to calibrate their frequencies to match the calculated Surry impact given in NUREG-1150. The four representative release scenarios were input to MACCS with parameters set identically to those used in the NUREG-1150 study. The NUREG-1150 analysis predicted a mean probability per reactor-year of 2.0 x 10-6 and 5.23 x 10-3 for prompt and latent fatalities, respectively. As shown in Table I, the use of the four input scenarios results in a good correlation to the NUREG-1150 calculated prompt fatality rate but over-predicts the latent fatality rate. This was accounted for through a reduction factor on the annual frequency of scenario 17-1. Reducing the frequency of this scenario to 1.4 x 10-5 per year results in a latent fatality rate matching that given in NUREG-1150 and still retains the balance of the source term groupings. A-1

FIGUREA-1 Distribution of Source Terms with Nonzero Early Fatality

• SUR-to SUR'45 SUR-11 Early Health Effect Chronic Health Effect Distribution of Source Terms with Zero Early Fatality Chronic Health Effect Source Groups: Representative Source Term; I I I 11-3 7-3 Ill~ 3-3 IV I 17-1 A-2

FIGUREA-2 Selection Of Category Scenarios NUREG.ll~O .ReprtSentative . Combiaed.

• E ~ Annual Risk Ca~U4.3 . Groups
• Source Tenn Fffi)ueney Prompt Latent I 10,11,5 11-3 3.24E-07 l.31E-06 9.46E-04 II 6,7,8,12,13,14 7-3 9.34E-07 4.?0E-07 l.31E-03 m 1,2,3,4,9 3-3 l.72E-06 2.70E-07 l.37E-03 IV 15,16,17 17-1 3,74E-05 0 4.23E-03 Total Predicted Annual Risk: 2.0SE-06 7 .85E-03 NUREG-1150 Results: 2.00E-06 5.23E-03 A-3

TABLE A-1 OPERATING UNIT SOURCE TERMS EVALUATED WITH UNIFORM POPULATION (500 PEOPLE/SQ MILE) OVERALL WBGHTED ANALYSIS Pop4.ll*flon* 0 1 ~ :*. '.,; '* .. *. IDiN. Jlldkll *OUl.;_:;J_ ti. ~ :...

                                                ~~~ !"~~,* "-tt            .>* . *:_ , ~~-~~~                                       .'t . ' ~ p . - ~ * . <        ~!'0oM CFti.irtv .-. R~~Ytf :: %.trts.r.a,ao.t*:

l~prrRl-'lri;'..; .;,-:'*.  : ~~~l11 *.': IRl~fper.ffi.tf lhd . (hit.  : ';* (1111} . , : fwly * . ,-' LlltHt , .,,,;.;,* . :, . t.~t*llt . ~ - . : . ~,,_. : CM~ *E~ Cht!Hrk Colwbfli.c . l!titlt * ¢11romc eutaim. 0.5 1.2 0.5S 4 24E-08 2.28E*08 8.48 0.16 0.03 0.19 0.38 0 06 0 4 6.77 0.97 6.74 I 2 16 0 89 2 25E-08 4 51 0 10 0 02 0 12 0.40 0 07 0.47 3 69 0 68 4 37 16 2.1 1.18 1.11E*08 1.91E*08 2 22 0.95 2 1 32 1.73 4.86E*09 1 43E-08 0 97 0 72 0 22 0 08 0 30 0.33 0 12 0.46 1 61 0 58 2 t9 32 4 2.28 3 13E-09 1 27E-08 0 63 0 64 48 5.6 3 25 5 14E*10 9 68E-09 0.10 0.48 0.16 0.09 0.26 0.33 0.18 0 60 0.82 0 43 1 25

          !I 6          8.1               4.43      3.82E*11      6 43E-09           0.01          0.32 8.1         11 3                6.24      5 09E*13      3 91E-09           0 00          0.20       0.43         0.36        0.79  0.22   0.18    0.40     0.29       0.24     0.54 11 3          16.1                 8 es     1.75E-14      2 40E-09           0 00          0.12 16 1          20 9              11.77                                                                  0.49        0 95        1.44   0.17  0 32    0 49     0 12        0 23    0 34 25 8          32.2              18.34                                                                  0 47        1.40        1.87   0.12  0.36    0 48     0.05        0 16    0.21 32 2          40.2              22.90                     8.89E*10                         0.04 40 2             41             24 86                                                                  0 04        0.16        0.20   0 09  0.32    0.40      0.03       0.10    0 13 48            49             29 69                     611E*10                           0.03       0 05        0.19        0.24   0.07  0.31    0 39 0 02            0 08    0 10 Nole: Tabte blanks Indicate that no data was gen£!ra1ed for that. distance interval and effect

TABLE A-2 OPERATING UNT SOURCE TEAMS EVALUATED WITH UNIFORM POPULATION (500 PEOPLE/SQ MILE) AND STRATIREO POPULATION

   ~*ff*' DfflnWit~n *'
    *. ~~ * :*

0.5 1 2

~f4* ~ -

• iirn1*
• 1 2 1 6

                                        ,..~     '.

(1111) O 59 0 89

                                                         ~'nnbln.d ~ - T*tms .:. ,* .....'/"\'-
                                                               ~'
                                                         .\WI, MAH
                                                                   ,0,59 0 89
                                                                          ..,_.,d  po Snat A.l'ff 0 05 0 05 0.01 0 01
                                                                                                  ~-PCIO
                                                           . ~ .:. : . :. ........ : :. . : *.:. ...........:..

0 7 unit.~--: 0 21 1.6 2 1 1 18 1 18 0.05 2.1 32 1 73 1 73 0 06 0.04 0 5 0.31 3 2 4 2 28 2 28 0 06 0 03 4 8 2 77 2 77 0.06 4 8 56 3.25 3 25 0.05 0 05 0.50 on 15 6 8 1 4.43 43 0 05 0 06

I 8 1 11 3 6 24 6 24 0 05 0 08 1.3£, Lil O. O 11.3 16.1 8 85 8 85 0.05 0 12 16.1 20 9 11 77 11 77 0 25 0 43 0.49 2 41 20 9 25 8 14 74 14 74 0 24 0 72 25 8 32 2 18.34 18.34 0 24 0 86 0.48 3.H 32 2 40 2 22 90 22 90 0 65 2 02 0.2 41 24.85 2 85 0 60 1 21 0 40 2 Si 41 42 25.42 25. 2 0 62 0.35 42 43 26 03 26.03 0.60 0.37 44 45 27.25 27 25 0.59 0.72 46 47 28.47 28.47 0 58 0.71 48 49 29.69 29.69 0.57 0 71 0 39 1.9.! Total Dose (man-rem per year) 8.54 13.13 Note: Table blanks Indicate that no data was generated for that distance and effect

TABLE A-3 0.33 MILE EXCLUSION ZONE WITH EVACUATION OVERALL WEIGHTED ANALYSIS PopuJilfon DisirlbUU ' M *. . *. ' , ' AnnOtW _,: ~ -...: . ~-

                                                                                                             ! ..f:'.r,-
                                                                                                                              . OQtM,n.*.      *.'* '.:,

In~ ritdlu~

   , {km)
                    . _oh ou~id. r~~I
                  . nun) fap'Yfnlff W***II~
                                 , cmn ' ' .. e,rty rr' Ind du.t*f-i.lW flat*
                                                            ~~-~r~  ,;:
                                                                        % <if Saf*,V ~~

Lat*"' . : E*rfY , l*t*"l

1rt.11q1* ~VJ)

                                                                                                    .... 1'.~mblMd .

(ff~~.,..~-

                                                                                                                                      -:i,.'. *t* . .
                                                                                                                                ~mblPd -_( '

r) .. AiH fodlvt~I OQM

                                                                                                                                                              ~ P4f ffi-VrJ 00111blntd 0.5            1.2         0.59     4.24E-08        2.28E-08        8.48         1.14                       0. 19        0.44               0.26         6.74 1.2            1.6         0.89     2.25E-OB                         4.51        0.00                       0.12         0.47               0.14         4.37 1.6            2.1         1.18     1.11E-08        1.91E-OB         2.22        0.95 2.1            3.2         1. 73    4.86E*09        1.43E-08         0.97        0.72                       0.30         0.45               0.38          2.19 3.2               4        2.28     3. t3E-09       1.27E-08         0.63        0.64

I 0\ 4.8 6.6 3.25 5.14E-10 9.58E-09 0.10 0.48 0.25 0.60 0.73 1.25 5.6 8.1 4.43 3.82E*11 6.43E-09 0.01 0.32 I 8.1 11.3 6.24 5.09E-13 3.91E*09 0.00 0.20 0.79 0.40 1.36 0.54 11.3 16.1 8.85 1.75E*14 2.40E-09 0.00 0.12 16.1 20.9 11.77 0.00 1.44 0.49 2.47 0 34 25.8 32.2 18.34 1.87 0.48 3. 19 0.21 32.2 40.2 22.90 8.89E*10 0.04 40.2 41 24.85 0.20 0.40 2.87 0.13 48 49 29.69 6.11E*10 0.03 0.24 0.39 1.92 0.10 Note: Table blanks Indicate that no data was generated for that distance interval and effect

TABLE A-4

• 0.2 MILE EXCLUSION ZONE WITH EVACUATION OVERALL WEIGHTED ANALYSIS

~op~lon Dl$f'1~tlon , ,,** . ' ~,¥f~f f*'-~11 !'- ' . ' ~~ ~~n t ~"~_-*:*_, i :** lndlvld~ Den, 111~ij~~-UI ou"kf,.ntdlal ro,* -* .. ,.'.' (F,i,ttty pcir R*-Yr) * ~- (if a*r*W*G~J (~ Pf!lf ffx*V!'f (~ ~ lli;y~f '. * : *. ~~ 'pw !l**lr) *

* ('u,no *      (km>         *:,,..,l*: : hrtv,            ' ..~~~ ..,:.    -~~,fy*.,- ** -..~'*"'    * *. ~, Co~b.Wd : ., *Comb!~* * * -Ar~*           ComblMd 0.3           1.2          0.55       4.87E-08      2.33E-08             9.74          1.17                   0.24     0.44           0.24       7.54 1.2           1.6          0.89       2.26E-08                           4.51                                 0.12     0.47           0.16       4.37 1.6           2.1          1.18       1.1 t E-08     1.91 E-08           2.22          0.95 2 1           3.2          1.73       4.86E*09       1.43E-08            0.97          0.72                   0.30     0.45           0.38       2.19 3 2             4          2.28       3.13E-09       1.28E-08            0.63          0.64 4 B           5.6          3.25       5.14E-10      9 SBE-09             0.10          0.48                   0.25     0.50           0.73       1.25 5.6           8.1          4.43       3.82E*11      6.43E-09             0.01          0.32 8.1          11.3          6.24       5.09E-13      3.91E-09             0.00          0.20                   0.79     0.40           1.35       d.54 11.3          16.1          8.85        1.75E*H       2.40E*09            0.00          0.12 16.1          20.9         11.77                                                                               1.44     0.49           2.4B       0.34 25.8          32.2         18.34                                                                               1.87     0.48           3.19       0.21 32.2          40.2         22.90                      9.49E*10                          0.05 40.2            41         24.85                                                                               0.20     0.40           2.87       0.13 48            49         29.69                      6.40E*10                          0.03                   0.24     0.39           1.91       0.10 Note: Table blanks Indicate that no data was generated for that distance interval and effecl

TABLE A-5 0.2 MILE EXCLUSION ZONE WITHOUT EVACUATION OVERALL WEIGHTED ANALYSIS PopuJeiiJm! ,t Pop. 0o.. Po,tJlft"!on Dl,kJbOUon f"tld* ntd!U* ln~~ldU.l ~*t*lltY ft*~t OUpldj. rtldlus P~ ~11'141: (F*J*~IY ~t R**Y'1 ,C. Of: \l*f*IY

                                                                                  .. :: Go.if
                                                                                               '      Aq~~

I~~ ~~vr* ., .'. Cff"l' pw

                                                                                                         . C:Ontblfttd lt**Yr:).

Combined lndfvfd~I Doie

                                                                                                                                                           . 1"'111 fMI' fl**'lr) com~lnad

{km) lkhl) fmtJ EarJy Ut*nl Elll'IY Lattnl ArH 0.3 0.5 0.26 1 67E-07 1.17E-07 33.35 5.85 0.12 0.98 0.26 31 58 0.5 1.2 0.59 1.2 1.6 0.89 3.93E-08 7.86 0.19 0.76 0.68 7.00 16 2.1 1.18 2.51E-08 3 31E-Oa 5.02 1.66 2.1 3.2 1.73 1.23E-OB 2.46 0.43 0.64 0.58 3.14 I 00 3.2 4 2.28 6.04E-09 1.62E-08 1.21 0.81 4.8 5.6 3.25 1.45E-09 1.01E-OB 0.29 0.50 0.28 0.57 0.93 1.42 I 5.6 8.1 4.43 4.45E-10 6.59E-09 0.09 0.33 0.34 8.1 11.3 6.24 1.86E-11 3.86E-09 0.00 0.19 0.87 0.45 0.60 11.3 16.1 8.86 2.84E-09 0.14 16.1 20.9 11.77 1.44 0.49 0.72 0.34 25.8 32.2 18.34 1.87 0.48 3.18 0.21 32.2 40.2 22.90 9.49E-10 0.06 1.09 40.2 41 24.85 0.20 0.40 0.13 48 49 29.69 6.40E-10 0.03 0.24 0.39 1.32 0.10 Note: Table blanks Indicate that no data was generated for that distance Interval and effect

Enclosure 5 NUMARC ANSWERS 1V NRC QUESTIONS NRC Notice of Proposed Rulemaldng: Reactor Siting Criteria (57 Fed. Reg. 47802, Section XI-B - October 20, 1992) Question Number 1 In making use of both deterministic and probabilistic evaluations, how should they be combined or weighted; that is, should one dominate the other? (The NRC staff feels strongly that deterministic investigations and their use in the development and evaluation of the Safe Shutdown Earthquake Ground Motion should remain an important aspect of the siting regulations for nuclear power plants for the foreseeable future. The NRC staff also feels that probabilistic seismic hazard assessment methodologies have reached a level of maturity to warrant a specific role in siting regulations.) Answer Using parallel deterministic and probabilistic methods to determine the SSE ground motion at a site is fundamentally flawed, technically unimplementable and most likely will make stable regulatory decisions all but impossible to achieve. Fundamentally, there is currently no technical basis to reconcile the results of the deterministic and probabilistic evaluations. The issue is that any deterministic evaluation is simply one interpretation of a range of such alternative interpretations permitted by the data. Because seismic source interpretations are uncertain given today's knowledge, probabilistic evaluations must be the underpinning basis for determining the SSE at a site. Probabilistic methods generally accepted by the NRC and data should be an acceptable basis to perform a site-specific probabilistic seismic hazard analysis. The - analysis should be accompanied by site-specific geological and seismological investigations that are graded in detail with distance from the site such that adequate site-specific data are available to confirm the generally accepted seismic source interpretations. Question Number 2 In making use of the probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015, is the proposed procedure in Appendix C to DG-1015 adequate to determine controlling earthquakes from the probabilistic analysis? 1

Answer The probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015 are technically unsound. Substantive revision of Appendix C to DG-10_15 is required. Significant industry resources have been expended in addressing our concerns with the NRC staff and developing a modified DG-1015 in order to provide alternative approaches for the NRC staffs consideration. The industry proposal includes two additional appendices (E and F) to DG-1015 and warrants a conscientious review. Question Number 3 In determining the controlling earthquakes, should the median values of the \ seismic hazard analysis as described in Appendix C to Draft Regulatory Guide DG-1015 be used to the exclusion of other statistical measures, such as mean or 85th percentile? (The NRC staff has selected probability of exceedance levels associated with the median hazard analysis estimates as they provide more stable estimates of controlling earthquakes.) Answer Yes. It is well-known that median values are a more stable basis for decision making. With respect to establishing the reference probability for determining the SSE, the median hazard of the current population of plants is appropriate. However, for the purpose of determining the controlling earthquakes' mean magnitude and distance, it is industry's opinion that deaggregated mean hazard results should be used. Question Number 4 The proposed Appendix B to 10 CPR Part 100 states: "The annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median annual probability computed from the current [EFFECITVE DATE OF THE REGUIATION] population of nuclear power plants." This is a relative criterion without any specific numerical value of the annual probability of exceedance because of the current status of the probabilistic seismic hazard analysis. However, this requirement ensures that the design levels at new sites will be comparable to those at many existing sites, particularly more recently licensed sites. Method-dependent annual probabilities or target levels (e.g., lE-4 for Lawrence Livermore National Laboratory or 3E-5 for the Electric Power Research Institute) are identified in the proposed regulatory guide. Sensitivity studies addressing the effects of different target probabilities are discussed in the Bernreuter to Murphy letter report. Comments are solicited as to (a) whether the above criterion, as stated, needs to be included in the regulation and (b) if not, should it be included in the regulation in a different form (e.g., a specific numerical value, a level other than the median annual probability computed for the current plants)? 2

Answer Full elaboration of this criterion, as stated in the question, places current technology (the LL.NL and EPRI methodologies) in the regulation as a basis for decision making. Clearly, these methodologies ,will be superseded by inevitable technology improvements. The requirement therefore, should be reduced to the foµowing: "The annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptable if it is less than or equal to the median annual probability of exceeding the SSE of the current [EFFECTIVE DATE OF THE FINAL RULE] population of nuclear power plants." Further elaboration of this statement would only lead to difficulty in application, and is not needed. Therefore, the additional criterion suggested above should not be included in the rule. Question Number s - For the probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power plants? (For the four trial plants used to develop the criteria presented in Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probability of exceedance level. Controlling earthquakes were evaluated for this frequency band, for the average of 1 and 25 Hz spectral responses, and for peak ground acceleration.) Answer The Safe Shutdown Earthquake Ground Motion response spectrum should be determined based on scaling an accepted response spectrum shape to the probabilistic seismic hazard analysis results for the average of 5 Hz and 10 Hz and the average of 1 and 2.5 Hz spectral accelerations, consistent with the reference probability level Magnitude-distance pairs for these spectral accelerations should be used to determine the acceptable response spectrum shape. 3

NUMARC Comments March 18, 1993 Line In/Line Out Enclosure 6 1 APPENDIX B TO PART 100 -- CRITERIA FOR THE SEISMIC AND GEOLOGIC SITING OF NUCLEAR 2 POWER PLANTS ON OR AFTER [EFFECTIVE DATE OF THE FINAL RULE] 3 4 General Information 6 This appendix applies to applicants whe apply for an early site permit or 7 combined license pursuant to Part 52 of this chapter, or a construction permit 8 or operating license pursuant to Part 50 of this chapter on or after [EFFECTIVE 9 DATE OF THE FINAL RULE]. However, if the construction permit was issued prior 10 to [EFFECTIVE DATE OF THE FINAL RULE], the operating license applicant shall 11 comply with the seismic and geologic siting criteria in Appendix A to Part 100 12 of this chapter. 14 I. Purpose 15 16 General Design Criterion 2 of Appendix A to Part 50 of this chapter 17 requires that nuclear power plant structures, systems, and components important 18 to safety be designed to withstand the effects of natural phenomena such as 19 earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of 20 capability to perform their safety functions. It is the purpose of these criteria 21 to set forth the principal seismic and geologic considerations that guide the 22 Commission in it& evaluation of the suitability of proposed sites for nuclear 23 power plants and the suitability of the plant design bases established in 1

NUMARC Comments March 18, 1993 Line In/Line Out 1 consideration of the seismic and geologic characteristics of the proposed sites. 1 2 3 These criteria are '3ased eR the Cl:irreAt gee13hysi cal, geel egi cal, aAd 4 sei smel egi cal i AfermatieA ceRcerni Ag fal:ilts aAd earthql:iake ecc1:JrreAces aAEi 5 effects. They *n1i 11 '3e revised as Aecessary wheA mere cem13l ete i Afenftati eA 6 '3ecemes availa'3le. 7 II. Scope 9 10 These criteria, which apply to nuclear power plants, describe the nature 11 of the investigations required to obtain the geologic and seismic data necessary 12 to determine site suitability and provide reasonable assurance that a nuclear 13 power plant can be constructed and operated at a proposed site without undue risk 14 to the health and safety of the public. Geologic and seismic factors required to 15 be taken into account in the siting and design of nuc l e~r power pl ants are identified. e6 17 The investigations described in this appendix are within the scope of 18 investigations permitted bys 50.l0(c)(l) of this chapter. 19 Each applicant for a construction permit, operating license, early site 20 permit, or combined license shall investigate all seismic and geologic factors 21 that may affect the design and operation of the proposed nuclear power plant 22 irrespective of whether such factors are explicitly included in these criteria. 23 Beth aetermi Ri sti c a Ad 13rebabil i sti c e Evaluations must be conducted to determine 24 site suitability and values- of seismic design parameters req1:JiremeAts fer the 25 26 si seuso, eA ape ~Pev1 dee 1A Pegul atepy gu1 dee aAd etaAdaFd Pevi e;1 ~l aA eeet1 eAe, 27 2

NUMARC Comments March 1a, 1m Line In/Line Out 1 -s-4-te. AdditioRal iRvestigatioRs or more coRservative determiRatieRs thaR these 2 iAcluded iFI Uiese criteria may be required for sites locateq iFI al"eas with 3 complex geology, receF1t tectoF1ic deformatioFI, or iFI areas of high seismicity. If 4 an applicant believes that the particular seismic and geologic characteristics 5 of a site indicate that some of these criteria, or portions thereof, need not be 6 satisfied, the applicant shall identify the specific sections of these criteria 7 in the license application and present supporting data to clearly justify such departures. The Di rector, Offi cc of Nuclear Reactor Regulation approves any 9 deviations. 10 11 III. Definitions 12 13 As used in these criteria: 14 A capable tectoAic source is a tectoRic structure that caFI geF1erate both 15 earthquakes aFld tectoF1ic surface deformatioFI such as faultiFlg or foldiFlg at or F1ear the surface iFI the preseFlt seismotectoF1ic regime. It is characterized by 17 at least OFle of the followiRg characteristics: 18 (1) The preseF1ce of surface or F1ear surface deformatioFI of laFldfems er 19 geologic deposits of recurriAg F1ature *t~ithiFI the last approximately 600,000 years 20 er at least 0F1ce iFI the last approximately 60,000 years. 21 (2) A reasoF1abl e asseci ati OFI with OF1e er more large earthquakes er 22 sustaiF1ed earthquake activity that is usually accompaFlied by sigF1ificaF1t surface 23 deformatieFI. 24 (a) A structural asseci ati OFI \rd th a capable tecteFli c source havi Fig 25 characteristics iFI paragraph (Ill) (1) of this defiRitioFI so that movemeFlt OFI OFle 26 could be reasoF1ably expected to be accompaF1ied by movemeRt OFI the other, 3

NUMAFCO:ll'nment8 March 18, 1993 Une In/line Out 1 In same cases, tl:ie geel ogi c evidence ef 13ast act hi ty at er Aear the gre1:1Ad 2 surface al eAg a particular cai:,abl e tecteAi c source 1J1ay be ehscured at a 3 particular site. This might accur, fer mcample, at a site ha¥ing a Eleep 4 everbureen. Fer these cases, evi eence may exist el se~1here al eng the struct1:1:Pe 5 frem which an evaluation of its characteristics in the vicinity ef the s1te caA 6 be :Peasenably easee. This e¥ieence 1Rust be used in eletermining whether the 7 structu:Pe 1s a capable tectonic so1:1rce within this definitien. ~ Notwithstanding i:,aragraph (1), (2) and (3) ef this definition, structur-a+ 9 assaciatien ef a structure with gbolegic structural features that are 10 geologically ale {at least pre Q1:1aternary) s1:1ch as many of these fouAEI in the 11 E:astern regien af the United States must, in the aesence ef coeflicting e*t'ieence, 12 demenstrate that the str1:1cture is net a capable tectonic source within this 13 defi ni ti eA. 14 15 ~2mbiaed liceasc means a cembiAee coF1str1:1ction permit aAd operating license 4'5 with eeAditions fer a n~wlear power facility issued pursuaAt ta Subpart C ef Part 17 62 af this chapter. 18 19 A deterministic source earthguake (DSE) is the largest earthquake that can 20 reasoAably be expectee to ~~cur in a given seismic source in the current tectenic 21 regime, anEI is ta be used in a determiAistic analysis. It is generally baseEI en 22 the maximum historical earthquake associated with that seismic seurce, unless 23 recent geological evidence warrants a larger earthquake, er where the rate ef 24 eccurrence of earthquakes indicates the likelihood ef larger than the largest 25 historieal event. 26 4

NUMAFC Comments March 18, 1993 Line lnfUne Out 1 Earl 'I Sjte Permit meaRs a Ge111Ri ssioR appreval, i ss1:Jet:I p1:Jrs1:JaRt ta s1:Jbpart 2 A ef Part 62 ef this chapter, fer a site er sites fer eRe er mere Rl:Jclear pe~~er 3 facilities. 4 5 A llill is a tecteRic str1:Jct1:Jre aleRg which differeRtial slippage ef the 6 adjacent earth materials has ecc1:Jrred parallel te the fract1:Jre plaRe. A faYlt 7 may have ge1:Jge er breccia beb.'eeR its b e walls and incl1:Jdes any asseciated 1 ~ meRecliRal flex1:Jre er ether similar geelegic str1:Jct1:Jral feat1:Jre. 9 10 The maqAitMde ef aR earthquake is a meas1:Jre of the size of aR earth~Yake 11 aRd is related to the eRcirgy released iR the form of seismic waves. HagRitYde 12 meaRs the Rl:Jmerical value oR a staRdardized scale sYch as, b1:Jt Rot limited to, 13 MomeRt MagR itYde, SYrface Wave MagR itude, Body Wa 11e MagR it1:Jde, or Richter 14 HagRit1:Jde scales. 15 e A response spectrum is a plot of the maximum responses- (acceleration, 17 velocity, &r- ~nd displacement} of a family of idealized single-degree-of-freedom, 18 oscillators as a function of the natural frequencies of the oscillators for a 19 given damping value. The response spectrum is calculated for a specified 20 vibratory motion input at the oscillators' supports. 21 22 The Safe Shutdown Earthquake Ground Motjon (SSE} is the vibratory ground 23 motion for which certain structures, systems, and components must be designed to 24 remain functional. 25 26 A scjsmic soijrce is a geReral term referriRg to both seismogeRic soYrces 5

NUMARCCanments March 18, 1993 Une ln/Une Out 1 and eapahle teetenic *seurces. 2 3 A seismegenie seurce is a pertieA ef the earth that has uniferm earthquake 4 peteAtial (same determiAistie seuree earthquake aAd frequeAey ef reeurreAee) 5 distiAct frem the surreuAdiRg area. A seismegeRie seurce will Rat cause surfaee 6 displacemeRts. SeismegeRic seurces caver a ~1ide raRge ef pessihilities frem a 7 well defiRed tecteRie strueture te simply a large regieR ef diffuse seismicity ea (seismeteeteRiC pre*,ince) theught te he characterized h:Y the same earthquake 9 recurreAee medel. A seismegeAie seuree is alse charaeterized l:J:Y its invelvement 10 iR the eurreRt tecteRic regime as reflected iR the Quaternar:Y (appreMimatel:Y the 11 last 2 million :)'ears) geelegic histery. 12 13 Surface deformation is distortion of soils or rocks at or near the ground 14 surface by the processes of folding, ir faulting, cempressien, er eMtensieA as 15 a result of various earth forces. Tectonic surface deformation is associated es with earthquake processes. 17 18 Surface faulti AEJ is differential greuAe ei spl acement at er near the surfaee 19 eaused di rectl:Y h:Y fault mevement aAe is ei sti Act frem RentecteAi c ty13es ef 20 greuRd disruptieRs, such as lanesliees, fissures, ane craters. 21 22 23 24 IV . Required Investigations 25 26 The geological, seismological, and engineering characteristics of a site 6

NUMARC C,omments March 18, 1993 Line In/Line Out 1 and its environs must be investigated in sufficient scope and detail to permit 2 an adequate eval uation of the proposed site, to provide sufficient information 3 to support beth prebabilistic and deterministic e¥a1Yatiens reqYires by these 4 criteria R!91Ri!l.:::~:~:y1:l::11:1:i:1l::::*:11:!::1m:~:1:::::::l11:,,:1ni::111i111r1, and to permit adequate 5 engineering solutions to actual or potential geologic and seismic effects at the 6 proposed site. The size of the region to be investigated and the type of data 7 pertinent to the investigations must be determined by the nature of the itel:i1J:e ea iiil!}::1:1J!n:iti!inlfi1i region surrounding the proposed site. The investigations must 9 be carried out by a review of the pertinent literature and field investigations 10 as identified in paragraphs IV(a) through (e) of this appendix. 11 (a) Vibratory Ground Motion . 12 The purpose of these investigations is to obtain information needed to 13 assess the SSE Safe ShYtdewn EarthqYake greYAd meti en. The seismic seyrces 14 (capable tectenic seYrces and seismegenic seYrces) in the site regien mYst be 15 i sent i fies and e*tal Yates. The setermi Ai st i c seYrce earthqYakes must be e¥al Yates -6 fer each seismic seyrce. 17 (b) Tectonic Surface Deformation . 18 The purpose of these investigations is to assess the potential for tectonic 19 surface deformation near the site and, if any, 1111:mJli:i to what extent the 20 nuclear power plant needs to be designed for these occurrences. 21 (c) Nontectonic lit!!F! Deformation. 22 The purpose of these investigations is to assess the potential for surface 23 deformations not directly attributable to tectonic-s- fiiiili, such as those 24 associated with subsidence or collapse as in karst terrain, glacially induced 25 offsets, and growth faulting. Paragraph IV(b) cencerns in¥estigatiens reqYires 26 fer tectenic sYrface sefermatien that can eccYr ceseismically. Nontectonic 7

NUMARCComments March 18, 1993 Une ln/Une Out 1 phenomena can represent significant surface displacement hazards to a site, but 2 can in many cases be monitored, controlled, or mitigated by engineering, or it 3 can be demonstrated that conditions that were the cause of the displacements no 4 longer exist. Geological and geophysical investigations must be carried out to 5 identify and define nontectonic deformation features and, where possible, 6 distinguish them from tectonic surface displacements. If such distinction is not 7 possible, the questionable features must be treated as tectonic deformation. (d) Seismically Induced Floods and Water Waves. 9 The purpose of these investigations is to assess the potential for nearby 10 and distant tsunamis and other waves that could affect coastal sites. Included 11 in this assessment is the determination of the potential for slides of earth 12 material that could generate waves. Information regarding distant and locally 13 generated waves or tsunamis that have affected the site, and available evidence 14 of runup and drawdown associated with these events, shall be analyzed. Local 15 features of coastal or undersea topography which could modify wave runup or drawdown must be considered. For sites located near lakes or rivers, analyses 17 must include the potential for seismically induced floods or water waves, as, for 18 example, from the failure during an earthquake of a dam upstream or from slides 19 of earth or debris into a nearby lake. 20 21 The purpose of these investigations is to assess the potential volcanic 22 hazards that would adversely affect the site. 23 24 V. Seismic and Geologic Design Bases 25 26 (a) Dete,minatien ef Dete,ministie Seu,ee Ea,th~uakes. 8

NUMARCCommen1s March 18, 1993 Une ln/Une Out 1 fer each seismegeRic aRd capable tecteRic seurce ideRtified iR Paragraph 2 IV(a), the determiRistic seurce earthquake must be e¥aluated. A~ a miRimum, the 3 determiRistic seYFCe eaFthqYake mYst be the laFgest histeFical eaFthqyake iR each 4 seYf'Ce. The YRceFtaiRty iR determiRiRg the deteFmiRistic seYrce eaFthqYakes myst 5 be acceYRted fer iR the prebabilistic aRalysis. 6 (b) Determ;natien ef the Greund Metien at the Site. 7 The greYRd metieR at the site mYst be estimated fr,em all eaFthquakes, ea iRclYdiRg the detcrmiRistic seYrcc earthquake asseciatcd with each seyrcc, which 9 ceYld petcRtially affect the site YsiRg beth prebabilistie aRd deteFmiRistic 10 appreaches. IA the determiRistic appreach, the determiRistic SBYFce eaFthqyake 11 asseciated with each seYrce mYst be assumed te eccYr at the part ef the seYFce 12 which is clesest te the site. AppFepFiate medels, iRclYdiRg lecal site 13 eeRditieRs, mYst be ysed te acceuRt feF uRcertaiRty iR estimatiRg the gFeYRd 14 meti eR feF the site. The greuRd meti eR is defi Red by beth heFileRtal aRd 15 ¥ertical free field greuRd metieR respeRse spectra at the fFee gFeYnd syrfaee eF ~ hypethetieal Feck euterep, as apprepriate. 17 W :f:im Determination of Safe Shutdown Earthquake Ground Motion. 18 The §§!~ Safe ShutdewR Earthquake GreuRd Meti eR is characterized by free-19 field ground motion response spectra at the free ground surfacei eF hypethetical 20 Feck eYterep, as appFepriate. These spectra arc dc¥cleped frem er cempared te 21 the greuRd metieRs determiRcd iR Paragraph V(b). Deterministic aRd p 22 lrobabilistic seismic hazard evaluations must be iiilPiil!i:):lffigi::ililiifdJliJiifhi!I:111 23 Ysed te assess the adequacy ef the Safe ShutdewR Earthquake Greund MetieR . The 24 annual probability of exceed i ng the SSE Safe Shutde~*R Earthquake GreuRd MetieR 25 is considered i1~iel!i§~i~ acceptably lew if it is less than ir:::::1111.m::::11 the median 26 annual probability of ii~iil:i::ni::::::~~:1:::::1§:1::::::e:r cemputed frem the current [EFFECTIVE 9

NUMAACComments March 18, 1993 Line ln/Une Out 1 DATE OF THE FINAL RULE] population of nuclear power plants. 2 At a minimum, the herizental Safe Shutdewn farthquake Greund Hetien at the 3 feundatien le¥el ef the struetures must he an apprepriate respense speetrum with 4 a peak greund aeeeleratien ef at least O.lg. 5 ilillllli:!1:Ill!Jlnlfi!il§iilllifiii:1:i!ililIIi!ffl:I:iinimiil::::::iniii:lli!:1:::::1:11111n11:1:1n 6 iiiieiii[l:iiiiti:riI::::11:11:iiII!ii@fliiiPl!IIi:irf:iiilfi 7 (e I,) Determination of Need To Design for Surface Tectonic and Nontectonic

• Deformations.

9 Sufficient geological, seismological, and geophysical data must be provided 10 to clearly establish that surface deformation need not be taken into account in 11 the design of a nuclear power plant. When surface deformation is likely, an 12 assessment of the extent and nature of surface deformations must be 13 characterized. 14 (e q) Determination of Design Bases for Seismically Induced Floods and 15 Water Waves. -617 The size of seismically induced floods and water waves that could affect a site from either locally or distantly generated seismic activity must be 18 determined, taking into consideration the results of the investigation required 19 by paragraph IV(d) of this Appendix. 20 (f !,) Determination of Other Design Conditions. 21 (1) Soil Stability. ~ihratery greund metiens determined in Paragraph ~(h) 22 ean eause sail instahility frem greund disruptien sueh as fissuring, lateral 23 spreads, differential settlement, and liquefaetien, whieh is net direetly related 24 te surfaee faulting. Geological features that could affect the foundations of the 25 proposed nuclear power plant structures must be evaluated, taking into account 26 the information concerning the physical properties of materials underlying the 10

NUMAAC Qimments March 18, 1993 Une ln/Une Out 1 si te and the effects of the §§l;!i!i!lii]!~iji,;;!;il:!:I@;!~:: vil:>ratery greYAd 1RetieA deteFIRiAed 2 iA PaFagFaph ¥(1:>). 3 (2) Slope stabili ty. Stability of al l slopes, both natural and artificial, 4 IRYSt l:>e eeAsideFed, the failure of which could adversely affect the nuclear power 5 plant . An assessment must be made ef the peteAtial effeets ef eFesieA eF 6 depesitieA aAd ef eelRl:>iAatieAs ef eFesieA BF EtepesitieA with seis1Rie aetivity, 7 taking into account information concerning the physical properties of the ~ materials underlyi ng the si te and the effects of the lll,:::::~qi,[l,~i,I:1:l!li!t:: vil:>rateFY 9 gFBYAEI IR8t i 8A deteFIRi AeEI i A Paragraph 'I ( I:>)

• IPl1n1:~:1l:::::llliusm~i1m:::I1£~isi1::I::1~

1o 111si:1ni:1111:i11i:111,:~:11r11i: 11 (3) Cooling water supply . Assurance of an adequate cool i ng water supply 12 for emergency and long-term shutdown decay heat removal shall be considered in 13 the design of the nuclear power plant, taki ng into account information concerning 14 the physical properties of the materials underlying the site, the effects of the 15 1:§~ Safe ShYtdewA farU1(:IYake GreYAEI MotioA , and the des i gn basis for tectonic and e, nontectonic surface deformation . Considerat i on of river blockage or diversion or 17 other fa i 1ures that may block the fl ow of cooling water, coasta 1 up 1i ft or 18 subs i dence, tsunami run up and drawdown, and the fa il ure of dams and intake 19 structures must be included in the evaluation where appropriate . 20 ( 4) Distant structures . Those structures that are not located in the 21 i1R1Rediate vieiAity ef the site but are safety- related must be des i gned to 22 withstand the effecti of the 111::::1:n:~:::::ePlin1:~::1l:::::::,:11:1121Ili:liiiil:~:1,m Safe ShYtdewA 23 faFth(:IYake GreYAEI MetieA. The desigA l:>asis fer syrfaee fayltiAg IRYSt l:>e 24 EteteFIRiAed eA a l:>asis ee1Rparal:>le te that ef the AYelear pewer plaAt, takiAg iAte 25 aeeeYAt the 1Raterial YAElerlyiAg the strYetYres aAEI the EtiffereAt leeatieA with 26 Fespeet te that ef the site. 11

NUMARCComments March 18, 1993 Line ln/Une Out 1 VI. Application To Engineering Design 2 3 4 Pursuant to the seismic and geologic design basis requirements of 5 paragraphs V(a) through (f I,), applications to engineering design are contained 6 in Appendix S to Part 50 of this chapter for the following areas: 7 (a) Vibratory ground motion. (1) Safe Shutdown Earthquake Ground Motion (SSE). 9 (2) Operating Basis Earthquake Ground Motion (QBE). 10 (3) Required Plant Shutdown. 11 (4) Required Seismic Instrumentation. 12 (b) Surface Deformation. 13 (c) Seismically Induced Floods and Water Waves and Other Design 14 Conditions. 15

• 17 18 12

NUMARCO>mments March 18, 1993 Line In/Line Out Enclosure 7 1 APPENDIX S TO PART 50 - EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS 2 3 General Information -45 This appendix applies to applicants whe apply for a design certification 6 or combined license pursuant to Part 52 of this chapter or a construction permit 7 or operating license pursuant to Part 50 of this chapter on or after [EFFECTIVE 8 DATE OF THE FINAL RULE]. However, if the construction permit was issued prior 9 to [EFFECTIVE DATE OF THE FINAL RULE], the operating license applicant shall 10 comply with the earthquake engineering criteria in Section VI of Appendix A to 11 10 CFR Part 100. 12 ~ I. Introduction 14 15 Each applicant for a construction permit, operating license, design 16 certification, or combined license is required by §50.34(a)(l2), §50.34(b)(10), 17 and General Design Criterion 2 of Appendix A to this Part to design nuclear power 18 plant structures, systems, and components important to safety to withstand the 19 effects of natural phenomena, such as earthquakes, without loss of capability to 20 perform their safety functions. Also, a condition of -a++ operating licenses for 21 nuclear power plants i!fflirtilli:;IIIrii.:11::11:~Jjij, as specified in §50.54(ee), is plant 22 shutdown if the criteri a in Paragraph IV(a)(3) of this appendix are exceeded. 23 These criteria implement General Design Criterion 2 insofar as it requires 1

NUMARCComments March 18, 1993 Une In/Line Out I structures, systems, and components important to safety to withstand the effects 2 of earthquakes. 3 4 II. Scope 5 6 The evaluations described in this appendix are within the scope of

• 8 9

investigations permitted by §50.l0(c)(l) of this chapter

• III. Definitions 10 11 As used in these criteria:

12 13 Gembi Ae<:1 li ceAse meaAs a cembi Ae<:1 ceAstFYcti eA peFmi t aA<:1 epeFati Ag 1i ceAse 14 with ceA<:iitieAs feF a AYcleaF pe~eF facility issYe<:1 pYPSYaAt te Subpa,t C ef Pa,t 15 §2 ef this ehapteF. 17 QesiaA CeFtificatieA meaAs a CemmissieA appFe¥al, issue<:1 puFsuaAt te 18 SYbpart 8 ef PaFt §2 ef this ehapter, ef a staA<:tar<:1 desigA fer a AYcleaF peweF 19 facility. A <:tesigA se appFe¥e<:1 may be FeferFe<:1 te as a "eertified staAdar<:1 20 f:lesigA." 21 22 The Operating Basis Earthquake Ground Motion {OBE) is the vibratory ground 23 mot 1on fe, which these featYPes ef the AYcl ea, pewer pl aAt Recessary fe, 24 ceAtiAue<:1 epeFatieA witheYt YA<:iue risk te the health aA<:1 safety ef the public 25 will remaiA fuAetieAal. The QperatiAg Basis Earthquake GreuA<:1 MetieA is eAly 26 associated with plant shutdown and inspection unless specifically selected by the 2

NUMARC C,omments March 18, 1993 Line In/Line Out 1 applicant as a design input. 2 3 A response spectrum is a plot of the maximum responses (acceleration, 4 velocity, &Fill displacement) ef a family of idealized single-degree-of-freedom 5 oscillators as a function of the natural frequencies of the oscillators for a 6 given damping value. The response spectrum is calculated for a specified 7 vibratory motion input at the oscillators' supports. -89 The Safe Shutdown Earthquake Ground Motion (SSE) is the vibratory ground 10 motion for which certain structures, systems, and components must be designed to 11 remain functional. 12 13 The structures, systems, and components required to withstand the effects 14 of the sq Safe ShutaewA ~arthauake GreuAS MetieA or surface deformation are 15 those necessary to assure: (1) The integrity of the reactor coolant pressure boundary, 17 (2) The capability to shut down the reactor and maintain it in a safe 18 shutdown condition, or 19 (3) The capability to prevent or mitigate the consequences of accidents 20 that could result in potential offsite exposures comparable to the guideline 21 exposures of §50.34(a)(l) of this chapter. 22 23 Surface deformation is distortion of -s&-i-1-s- ii:illiiiliiiilnilit or rocks at or 24 near the ground surface by the processes of foldingT ~~ faulting, eempressieA, 25 er exteAsieA as a result of various earth forces. Tectonic surface deformation 26 is associated with earthquake processes. 3

NUMARC Comments March 18, 1993 Une In/Un* Out 1 IV . Appl icat ion To Engineering Des ign 2 3 The following are pursuant to the seismic and geolog ic design basis 4 requirements of Paragraphs V(a) through (f I,) of Appendix B to Part 100 of this 5 chapter: 6 (a) Vibratory Ground Motion. -87 9 (1) Safe Shutdown Earthquake Ground Motion. Ea,thquake G,euAd Met i eA must be characterized by free-field ground motion response spectra at the free ground surface e, hypethetieal ,eek eute,ep, as 10 app,ep,iate . IA view ef the limited data availahle eA vih,ate,y g,euAd metieAs 11 ef st,eAg ea,thquakes, it usually will he app,ep,iate that t lhe design response 12 spectra f,b.9#1::1 be smoot hee spectra fiJll:l=mt:1k1=11:01.$q11,pgpq0,l1lb.tnemif1¢.t'i:¥~M:M:l:P!n~:@ 13 iMI11i£iU.iilij'.i$!JEWi&it.iiiMn:irid:r::ijd.iii)i.i1:itiIJl,iliil,iil!Iii::[:::111,t1~: de¥ el ep ed ff'8RI aA 14 eASeRlhle ef f'espeAse speet,a ,elated te the vih,ate,y metieAs eaused hy Rl8f'e thaA l,liitn!nl::I;\§;filllt!iiii:il:1,§1 15 8Ae ea,thqYake. ~t a RliAiRIYRI, t ihe horizontal We ShutdewA Ea,thquake G,euAd Meti eA at 1:i the ill!iliil!Jil,ii!:iil,l,]111¥: foundation level 17 of the structures must be an appropriate response spectrum;:\Iiifi.l)lliriHFJ.llfil:::::1:i:ii. 18 li!l!!lili:iil[i::ll§lllt:ill:f:l with a peak ground acceleration of at least 0. lg . 19 The nucl ear power plant must be des igned so that, if the §1:§: Safe Shutde~*A 20 Ea,thquake G,euAd HetieA occurs, certain structures, systems, and components will 21 remain functional aAd withiA applieahle st,ess aAd defef'Rlatien liRlits . In 22 addition to sei smic loads, applicable concurrent normal operating , functional , 23 and accident-induced loads must be taken into account in the design of these 24 safety-rel ated structures , systems, and components. The design of the nuclear 25 power pl ant must al so take i nto account the poss i bl e effects of the Ill We 26 Shutde~*A Ea,thquake G,euAd HetieA on the facility foundations by ground 4

NUMARCComments March 18, 1993 Line ln/Une Out 1 disruption, such as fissuring, lateral spreads, differential settlement, 2 liquefaction, and landsliding, as required in Paragraph V(f I> Qf Appendix B to 3 Part 100 of this e !hapter. 4 The required safety functions of structures, systems, and components must 5 be assured during and after the occurrence of the ~~ll \'i ln*ateF:Y gPe1md R1eti eA 6 asseciated ~,ith the Saf'.e Sh1:1tdewA EaPtht:11:take 6Pe1:1Ad MetieA through design, -78 testi.~, or i~l!r::::ii1~:1111111:1:::i::1uiUi9i:j,J!i t:11:tal i fl eati 8A The evaluation must take into account soil-structure interaction effects methods. 9 and the expected duration of vibratory motion. It is permissible to design for 10 strain limits in excess of yield strain in seR1e ef'. these Iii safety-related 11 structures, systems, and components during the §~li Saf'.e Sh1:1tdewA EaPtht:11:take 12 6Pe1:1Ad MetieA and under the postulated concurrent loads, provided the necessary 13 safety functions are maintained. 14 (2) Operating Basis Earthquake Ground Motion. QI; QperatiAg Basis Eartht:11:take 6re1:1Ad MetieA must be characterized 15 (1) The by response spectra. The value of the QI; QperatiAg Basis Eartht:11:take 6re1:1Ad 17 MetieA must be set to one of the following choices: 18 (A) One-third or less of the II§ Saf'.e Sh1:1tdewA Eartht:11:take 6re1:1Ad MetieA. 19 The requirements associated with this QII QperatiAg Basis Eartht:11:take 6re1:1Ad 20 MetieA in Paragraph (a)(2)(1 )(B)(I) can be satisfied without the applicant 21 performing explicit response or design analyses, or 22 (B) A value greater than one-third of the 11£ Saf'.e Sh1:1tdewA Eartht:11:take 23 6re1:1Ad MetieA. Analysis and design must be performed to demonstrate that the 24 requirements associated with this HI QperatiAg Basis EaPtht:11:take 6Pe1:1Ad MetieA 25 in Paragraph (a)(2)(i)(B)(I) are satisfied. The design must take into accQunt 26 soil-structure interaction effects and the expected duration of vibratory ground 5

NUMAR: Comments March 18, 1993 Une ln/LJne Out 1 motion. 2 (I) When subjected to the effects of the gJ. Qperating B~sis Earthquake 3 Greund Metien in combination with normal operating loads, all structures, 4 systems, and components of the nuclear power plant necessary for continued 5 operation without undue risk to the health and safety of the public must remain 6 functional and!l]Ji!!:iiiitiiiili:i!:IH: within applicable stress and deformation -87 9 limits.

ri1:i:::::::1ir%:1m1nl:$.:liiiiiiin11::::::1.,::::::g::m:IYil:I:ll§hlri[~§e:I1Hl'::::1,1:1:::::a111rmi:B.@:tirI:111.

1:1:11ii:111i:::111:::::11111i:iiilIIlll]liill:rlibilil1ll!i:a:1::1:!::),::1:111111::::g:::::1:11111:11i]iiilitlillll.Jlll 1 1 1 1

10 (3) Required Plant Shutdown . If vibratory ground motion exceeding that 11 of the I~; Qperati ng Basis Earthquake Greund Meti en lliEiil:~::::::1,~j:l1,~1:1:1: or if 12 significant plant damage occurs, the licensee must shut down the nuclear power 13 pl ant. Prior to resuming operations, the 1 i censee must demonstrate to the 14 Commission that no functional damage @i:1l!:!!i::::a:1 has eeeurred te those features 15 necessary for continued operation without undue risk to the health and safety of 41 the public . 17 (4) Requfred Seismic Instrumentation . Suitable 11!9.iilli instrumentation 18 must be provided so that the seismic response of nuclear power plant features 19 :illii:liltti:~J!iieiiil:i:;:{l!III!i!:!ilif important to safety can be evaluated promptly 20 after an earthquake. 21 (b) Surface Deformation . The potential for surface deformation must be 22 taken into account in the design of the nuclear power pl ant~ by pre¥iding -:* liln 23 :l[f!lll!:l!lli:I[@Rli!l:l:11:i:Ilt!mlaYIIIIIIll!IR!Blii@II!!Il:1::::11111n11:111::1111::::i111:11J11m111 24 :1111::111tlIImililIIirlilli reasonable assurance that in the event of deformation' 25 G11ieenee ie 1,ein!I ee,e~e11ell in 9reJt Ae1111lehry G11ille 96 1917, "Pre Eartlil1111elte Plennin11 26 a111l IR1Rolliate Nwol ea, Po~cer Pl a11t 911ento, Poot ,a,tlil11wako Aot~ 0111." 6

                                    \

NUMAACC,omments March 18, 1993 Line In/Line Out 1 certain structures, systems, and components will remain functional. In addition 2 to surface deformation induced loads, the design of safety features must take 3 into account seismic loads, iAeluEiiAg aftersheeks, and applicable concurrent 4 functional and accident-induced loads. The design provisions for surface 5 deformation must be based on its postulated occurrence in any direction and 6 azimuth and under any part of the nuclear power plant, unless evidence indicates -78 this assumption is not appropriate, and must take into account the estimated rate at which the surface deformation may occur. 9 10 (c) Seismically Induced Floods and Water Waves and Other Design 11 Conditions. Seismically induced floods and water waves from either locally or 12 distantly generated seismic activity and other design conditions determined 13 pursuant to Paragraphs V(e §) and (~I.) of Appendix B to Part 100 of this chapter 14 must be taken into account in the design of the nuclear power plant so as to -15 prevent undue risk to the health and safety of the public. 7

NUMAACComments March 19, 1993 Line In/Line Out Enclosure 8

                                                                                                          .November 1992 Division 1 Task DG-1015

Contact:

A. J. Murphy (301)492-3860 DRAFT REGULATORY GUIDE DG-1015 IDENTIFICATION AND CHARACTERIZATION OF SEISMIC SOURCES, QHE: RM IN I sr I c sou Re E: E:ARTHQU~KE:s, AND 1:1~111:~:o.m:1::1J::::::g1r::::::11:F::1]:1:rmiaN 1111nµ111 GROUND MOTION 1 A. INTRODUCTION 2 3 Paragraph IV, "Required Investigations," of proposed Appendix B, 4 "Criteri a for the Seismic and Geologic Siting of Nuclear Power Plants on or 5 After [Effective Date of the Final Rule]," to 10 CFR Part 100, "Reactor Site 6 Criteria," requires investigations to assess the proposed site for (a) vibra-7 tory ground motion, (b) tectonic surface deformation, and (c) nontectonic e: 10 deformation. Paragraph V (in a 1.nli?i threugh El) of Proposed Appendix B to 10 CFR Part 100 requires the determination of (a) EleterllliAistie seuree eartheiuakes, (b) greuAEI metieAs at the site, (e) -s:lafe ~lhutdown e§arthquake 11 9lround lllfflot ion llllliJi and (Elg} the need to design for surface tectonic and 12 nontectonic deformations. 13 14 The purpose of this guide is to provide general guidance on acceptable 15 procedures to (1) identify and characterize seismic sources, (2) determine 16 EleteFRli Ai st i e seUFee eaFth EjU ake s (QS E:s) 1.11:::::::1:1:i1::::::1~u::111:,::::::P.ilirl.Mii[{il ).::ra.iv~n::ppJ!::1 17 'Ht!:1m1:11::~;1:1:rlit1::111rm111}~9IIl!i!l{tI!ttsJ::MllfflltI!\l!l!!ffl!It!l:i11ttl!IIIPI ,lft6 1 18 eeAtFel 1iAg eartheiuakes (GE:s) illitHltfiii1:~::py~i,i]i11:I:~miiii.si:::::i~:1::::::11:rPliiliRiilltlfii.l 19 l!P~:Hi!::::::1fi1::::::~§I, and (31*> EBRlf) are lilim.~::11::::::1~11::::::1:1:11:1:1:1:111[::11:1e:11n:[1::::::s1m:irljlij~q 20 t.!i the median seismic hazard level te that at operating plants. These 1

NUMAAC Comments March 19, 1993 Line In/Line Out 1 procedures are required by Appendix B to 10 CFR Part 100. 2 3 This guide contains several appendices. Appendix A contains a list of 4 definitions of pertinent terms. Appendix B describes the acceptance 5 criteri~fi for the annual probability of exceedance level for lb.illS.$1 5-a.fe 6 shYtdowA eaFthqYake gFOYAd motioAs. Appendix C discusses the determination of 7 !liII§ll:1:tI!H@li11a~s;11:1:1tttef:t:111:111:s:111:1.ifflt:11:1:11:11mt:1:,1::1111Ji@M!llRRm!illltlI::Ji 8 :1(Hliffli.lilfniil:rl:::::::H~f.i'rm11:lintllii~:i eoAtFol l i Ag eaFthqYakes fFom the -~ 11 12 f)FObabil 1sti e aAd Appendix o discusses the 1111:~::!1l:::::::11:1!81pji::1:1~:s:[1ii1lliilJl:i1m:~:: l!:~:111~:11:;,:1:!!llll'l!m.t:9~9:p§yl,iU;.:1:J: i nvest i gat ions~~ to :H111:1:a:::::::1n1 eh aFaeteF i ze se i sm i e SO YFe e s::~::::::::::::::f:p~::::::~:~:~~~:::::::,::,~~~:~~::::M::,::::::l:~'-:i:;;.,;~1,:~::::::,~ijl:~:~~~~~:~I:::li.!:PrW::::t:,~:~l)M~:,~ 1111::::::11:1:1::::::1:1::::::1111:::1,1:11::::::rliil:11:n:1:::IH!il:bi:nl::111::::::111:1:tllil!:§ii:l[Rl::ili:flii:lii:l::l:::tiiii 13 :1r,m::::::11,;::::::1,11,::v1:111::1:t1:m11u:o.:1m::::::1u1ifili.:R::::::1::::::1.1$.:,r:1:li.lI::1::::::pfliilliiiltil!li.finw1:D.::tt.li.i 14 ll!:I::i!!jt!¥!:!i!!,::11i:il:l:i::1mJ:9:f:l:l:9!:lil::Iilni{i:11ittS!!I:!:l:!!l.!!l!!!:l:!Il.:l:!ilh1:fi1.1:1::1i:l:9! 15 lilillti!llli:t:11:111:::::::1,,1:!iit::::1:i1:1:i:l:]:::l:ffiHtt:::::,,n:1:i:mil1!l:]l!:ili!llt!li1.l::m,:::::::ri1!l:t\1!:1:ei 16 IRPlni:;:1,:i::::l:j:j:;Jg:::::!:l:l::::::il :'::::l1~~::ijl !:t1e:::j:jpn,qvj1:11:j:j::jipJ,!:9.p!,t,:1:#:1:::::::!:i::::!!:l1:::::::ri111l::~j,:j:::jgfj:::j:1,:rr1 17 !:J:IIJ::1'1!!:!t:1:111:i:!l!lill!!:!]f!!i:!llmn1r:11:1m1m:11::t!!P:!tl!fl:llll:lbl:i!:!:flil!Bll!ti]!I!:! 18 IPR@:rl:1:1.::::::1::::::i1111,::i1:1:i:th1:::IPffli1iitri1::::::i!fI:li!1m1::n1t:1:i1:{§§lt:i!ll:rt~I:!i)ijpgn1:1 19 iliiilriml: -~ 22 23 Any information collection activities mentioned in this regulatory guide are contained as requirements in the proposed amendments to 10 CFR Part 50 that would provide the regulatory basis for this guide. The proposed amend-ments have been submitted to the Office of Management and Budget for clearance 24 that may be appropriate under the Paperwork Reduction Act. Such clearance, if 25 obtained, would also apply to any information collection activities mentioned 26 in this guide. 27 28 B. DISCUSSION 29 30 AppeAdix B FeqYiFes eoAsideFatioA of both pFobabilistie aAd deteFmiAis 31 tie appFoaehes to obtaiA site geologie aAd seismologie ehaFaeteFisties. The 32 approach required by Appendix A to 10 CFR Part 100 for determining the 5-a.fe 2

NUMARCc.omments March 19, 1993 Line In/Line Out 1 shYtdowA eaPthquake gPOYAd motioA (SSEt is deterministic, and thus does not 2 explicitly incorporate uncertaint i es about -the i.!i:Ii:l!lii~:fi seismic hazard into 3 the ground motion determination . BltMin!:1n,1:ml:mi1.1il<<:l[fi1m::l1ung:JJ1tltti!! 4 Billijr.ililfiiji,fii((l~fiilU:::11r;1,rmii.J!iiiE((lriiRiliil:tifiit@l.l.inl.;1.IJ.lJ.vlJi¥):J:::i1ni9,ffilli.tg 5 lllll)J.;i:i~.Yt::J:1:1::1;11m1.v111.11J:At1:liTJ:li.)1:l$1U:1.:::::::ff11111:1;,1m:ner1JAll:liJ.::l;J:t¢.mlnffltllMII 6 lii!l@PB:I101l!1llP!l::lit:!:t1:::::ffip§))iii!IiJ.::1.nitiii:IJ:1u:11:11i111g11111~l!li:il(fi!IIJBI. 7 1un1:mu::fflen:111.t.n11.n1.t.~rm11i:PrP.:ti:d!f:i@i:1,t,r4nMillit*11r,1tb.::;;1,,r::1::1.e.,m,:1i1.*rrm1.Knm 8 w.ii#.lt::tt1:1111im::IP.t:1:b.e.#S.llr::w.br,uiJt)P.irft.-rn:t.9l1ffl.tutrn1!:lb.J;:::1:i:1i+ii11l:ta:g -~ 11 12 l:ii!ll:J:111:1:11111111mr:J1!!:liti1l]IIII[l~ll::i111:i1,i111ririlll'.IIII,fl!:1m111111111111:11 1.-n,tn,;teu:rthat::rnWJH:J:~g11J:td.A~ttm:t:n:t:$;tJ:,:1rerme.1e.nN.1t;l&rt:t~f:@i$.gt:11tn:e.n$,11a1e.e1::1::aou. iiP.Bii@ti.l!n1:r11::11@::1J.:11:::::l.iii):ffj§l)i:Uitiijj:l,:Iliillffiis,;U:mi.tili:iiliillJiiJ@fif:lttt 1111111:in1:i,:1:i::i1r::11:1:1n1:!::11:11::i::1111en1111J:1,1::::::e1:::::1111:rs1:11:i::i]tt1,:,1!1:111:i1mit::111 13 1~~1,1~:1:,~1.::r#.~~@.=~::::::]~:~§,~~~~::~::,.::1,:~1@::m~~:'-:~M.~:~:::::::~~~:~r.9.i:::::1,~~:,;~'-:~:::::::1:~~~1::::::::1w1:~:'-:#.~~:::rn~ 14 :1!iffliii:rl::::::11::::::11.l!ri:1:,:::::::'=tt@J!::J!iniir1:11::11:~:i~::::::11:iii:~::11iltl~:!lr::::1:11~::::::~:linl!ji!li!iiil!!Ji;i::::::inl 15 Fl!!!F!!rI!iili!!:J]gfiffif@l)i!@st:!ll9!fttii!]:li{IlnP21i9Cf:lll:ilt:fflij:J.:i;t:Pl:ltf:lf!llmlr@!ilil:PP.;§) 16 il:::::11:~:'=m~:11:~::iill:i:e"nliliii~:~: GurreAt ppobabi 1i sti e sei smi e hazapd aAalyses 17 Pely hea¥ily OA eMpePt opiAioA, aAd siAee theiP Pesults ape dPi¥eA by the 18 tails of the pPohahility distributioAs, they Reed to he heAehMaPked h;Y simpleP 19 detePmi Ai st i e aAal ys 1s. ThePefope, the Pole of the l!!ill:1:1::::~:111:::::::1,,.inl::::::1~:11:::::: -~ 22 23 probabilistic analysi sl is to eAsuPe that the uncertainties ha¥e heeA iAeluded in the assessmeAt ifiiiii,,lim1:~11:~:11 of the site sei smi e hazapd::::::111:e}:§,:Iiiiiiii: 111:::::::1:s111:1:1H~~JHtm.b:::~:n11IB!!l.!@g:::::::i:n::~::;J;l@:11111mJ:n:111:ifit:i11Mtb:1t:111::~~ , aAd t);he ro 1e of the detePmi Ai st i e aAal ys i S i!~;il!tt1fi9.l.!:~:::/i!9.i~l!!t!B!!i~!il1Jiq[l9ip~!:29.!~:i!i]i;fii 24 !ji,JJ[irop}jgj'¢j)Jj::;::ri.f<<f::J:JggQpn(i£jJjO.iJ:ia)tt;l.$.:IJ.!gi1Jjfqpif i S to ensure that the Pesul taAt 25 desigA pPo¥ides pPoteetioA agaiAst a seeAaPio based oA histoPieal seismieity 26 aAd peeeAt geol ogi eal hi stoPy iil!ili:1::::1i.aiiili:::)iriil:'.ys,=i::,m11~:d§U.i.U:::ir.n:1tif.lliilitj 27 :~::nf:~nmm!::1::pn. 28 29 l~itf:l,ijf!IIUil'lPIIJ[t:!lf::n1:t:IJ:1:1mi:tIIl1ilirtII!Mil:111i:!ttim:1ittlliljI!i:I:11:111 3o m1mi::n=r1n:1:1n1 :r111:1iir1.:i11:1m1::1111:::::::i1111:1::1ut1::::I11.111111:n1.:1:11r:111111:1111:::::11::11I:::11.1 1 3l sll!ti!lff~il!:il:1:Pm,::::efiti!:1:1:1m:t:P::t!)!MISl!:t:i:B!l¥tl::rn1nlrn:g1s:I1I§!!lf:i!J!:lf9~1MI.Sllr!l!J]1!1 32 ;.:o.QJ:UiN:fiU.m~iii:::§iiffilt:r&:~:::rru.µv.U.r1:ti:11mi.n1.11a:1nmt:t1.1+J.ht1::r,1;:1n1rrwiiiiimt:1.+~~1:1v:1::1J 3

N~C,omments March 19, 1993 Line In/Line Out 1 ;r,1r;e1:rt:earthqua&:::::J,cc;ur:mnce;tmo.da1::1:n1;tm:::etc@t?:rAsrd::sr::1.1amta.t.i.t=f.i.r.:ms:1::te.:$1:=;urntHe. 2 s.c.Rwfal:t1rnat!ht*fmo.lel:~n]nus.1:ib1d(tntfiti*fllr=to.:r1ce.o.unt=:if~>>:NtlAs'1tunce:rt1:1=nt.;J:-.i:=t: 3 4 tttlirt-.,:b.li@IIfll.lUi:111a.r::wl:llt::::ibti':iiff@niJ:1t§l1;l@ttiaml@i.nlr.lslirti¥.llii>>.nr:qf 5 :~~:,-:~11;~:~::{~f:Y.:f:¢~~)::mt.:1::::;~~~:;:::~~¢~V.llj:p.:f~~¥i:~p.fflllq~~IIJ.ll(?:,a.:;:::'-MIJJU~1,,~::wm::MJ11l:~ti:f.O.~ 6 ilni:!lii::::::1,:;:11::::::1111liil:::::::19::::::111ilil:iiil§§ffillI:l!1in::u1::::::1111:::1:,i1:~:: 7 8 -~ 11 12 :im::t:l:JI}:::;,r:!!ilim]:ln!IIl!li]/RDm:ll@Rll*:nil~t:11pJ:RSllE!!:i::r:,:itlimPl!!llll!l::l~@IiB 13 iiiiiii:t:sim::::::1::111i:t*i:s.lillill:Ji:i:::::::1,11 14 15 16 17 ib1lliJ:lliilllilli!{:li1llil@@ilii:llliilliilili1:1111111:st:11111111iI:1li!:Ililinlilfitiiil 18 slinislln:i:111:111::imJ:s:;:::j9~91i:::::::1,1:r:1e!::::llilllil!ii:IIIB:l::::111I:!i:ill::iii!il:!lerilil:lm1:11:i1 1 1 19 miill!illilMlilfl!J:1:ilinll!iiiml:[illili!iim!!nl.lli[ei§li§J:1::i::11:1:1::::::111:i1i11:1e111 -~ 22 23

   )!!!!!illl11iI:J:ili:!lill:IIIII!iillirJ:is:ii111:111::1n1:1::JJ:21:l:nli!PE!!lll:em::::11r1111:g:t,:t

11:tiffiil:ii:tit::1,1::::::1111iti:tii1!l:!1iti:t lsllilirlllni::::::1:11::::::!ll!ii:Iiin::::::ieiilliil!I::1111:::::11111111:i1{:::::111::::::1:n111 24 :i:n1irirlllil:1:11!:lmil,illlglillilll§.ll:f.::1::::::::::1inm1t1.:11.1:;;!iJ!n!::;:llt!i:/::1m:irn:11n1::::::1,,:,m.1::si:1,11,1 25 ffil:!:lllllJ:111::::::1111:::::11111:11fflI::llil;jillilBit:11:11;1iifflllll:!leti:11IIlllell!lr.<<:::::::1111H1:1:

26 f:fffl:IMtl:J.::)fjfig:{l§llifil=islnl:si:minl:::J~iilirill:Jfgj,)IJ.llBi::lil §1RJ:}::i:r111f:l:i!J;fl:iiiil!i1n 1 27 ili]:ililliiiDIIisiiiilll[:fi.![ilfii[:ililli:l:::1::::11,1111::::::111111111i::1:11n11:1i11::::::11H1:::::m111pg;, 20 :,,1:::::r:,:,m,,:;*1::::::11:1u.~a11::::1:1:tt11@1:1111I:11~111,1:::::ffi1t1.111.tiH:11::::::11,m1.s.1:11t:t:11}:1.i~:twn:]~11.1 29 111:~m:::::::::u.1..n11~::11;:1:1:::::,11:1:1::110.1:::::1.:111;1::111,1.::1::::110::1r1.1i:i:111@;:~1111::::::1.:1111.111111:,1r:1111:1,11:J;i, 30 §1l;M!!!:J:1b1:rn~n1.~1.J:l~t:!:l[:Jin!:t::~:~:i1mii!Jll*J:s1~11n::11:1:1m1~:g:fP.l):::~::r1.ams.11Pli'.i:J:~!,"l 31 iltiiill:1:,11:1:i.n'l1.il@lli.ife.'xi[1.tlJ:ii[):::)ii!J[$m~IWiJ.iffl,i.I:11f:;]pln1.ill@n1::r"':=:::1.1n11e.ua:;:::::::: 32 il!!:IP~mll!I:PEI!~l:if::!~~l:!:!IIP:lll!I!Iil!l)l::i,~lrfflH!ffl:J:::ft"l!l!l]:J~n~n::§:i:i\1:r:i:1111!:11:1!:J,:gn, 4

NUMAFC Comments March 19, 1993 Une ln/Une Out 1 .,,:,:.:,prmdu.ce.d::fnewmJJ.at:arn!intr;t,ntat1,te.t.1t.m&o:sw,t.H1itt:atJttno1t1,ae.ftua1teJ.,y:,r1t:nto.rP.nrate.a. 2 ,nrttfier:ex,i=stttn.g:::fs.e1smi:cJ1,10.urte:sno.r.wtnej,fdfa1ahamenrs.=1M=t:Ytd,§.gasse=s.sment:;::t*tUc6 3 ¢onii=has.:::ron:::,,,a:ts.:i,te&1,e.c.:i:'li=cmtias.ls:6Jhe.tge11~r.a.1tJtcctit.1ntg;;,,frthe.,tex,1=stln9 4 ~t.ilS,.,J,t&so.utte.J.¥@1,,,;8te.1nri,lfflmJ,n:tii?,MJ.ld!1=1falH*nl=tA$farei.tttll!Hl9.l!l.$./'il 5 ~§lilifflli:i¥!i!=l:j~:,-11..t'i~!:~: 6 7 As part of the EPRI sfudy, an extensive earth science database for the 8 SCR was developed . In addition, both !liIIl,ltll:]it@:=iil:11!: studies developed 9 scientific interpretations of available information, and determined seismic ao 11 sources and source characterizations for the SCR (e .g. , earthquake occurrenc.e rates, est imates of maximum magnitude). 12 13 lttitititH~iiaci:hiet::,1:a~e:,u:*arg:t:n@:m,orii:io.li:tHet=Westarn::a11:s:i::ii:a:i¢.o.maraH1:e 14 i!llJ::§ffllt:i::ill:li11Il!f::tiiiJ.II!liP!~Iins.ltli!i!Iinltils:Htill:f::i!sIH!ili!iiiillll;Pi!lli!!I 15 :111t!ll!!i!l!!9.ttillt!!lli!I!!!!9!11:I!Sbllii!trtl!:1!!]::g1:1!IIJ:!J!!R@tlllBl!:¥l!t!l!l!!lli!&1:!¥ 16 iilimiiinii::Iiili::~::1:=:::::1PiilliilI1iil[t!1I:rn1111::::::11itiI:Iril.:teiiI:t1:111:i:1~:sIIl:11inlnilit:l:in:1 11 1~1:!l!:!~tl@:1:i:t;tii:~; 18 19 -~ 22 23 llllliIIilll!ltf!l:'li!iY.IJl!le!till:J!l'!llll:!Il11illt!lilmfI:ltll!lll!llll!llllill!I! re1m::::::11n11:~n::::;:l:li~ln]!~g::::::~:11:,:::1111:t n1::,::::::11,~1~:;,::::::::1:::1,;:1,:,:I1:1:::!:!li!l!:::l:iEi:::m,~ilffll i:l::~::,: 1 24 The area east of the Rocky HoYAtaiAs withiA the North .OaericaA Plate aAe 25 well away from the acti¥e ~late margiAs is eescrihee as the "stahle coAti 26 AeAtal regioA" (SGR). In the SCR, characterization of se i smic sources is more 27 problematic than in the active plate margin region because there is generally 28 no clear association between seismicity and known tectonic structures. ,J,:q 29 iiiiii!;ll +,jhe observed geologic structures were generated in response to 30 tectonic forces that no longer exist and that bear little correlation with 31 current tecton ic forces. Thus , more j udgment must be used than for act i ve 32 plate marg i n regions , and it is important to account for this uncertainty by 5

NUMARCQ>mments March 19, 1993 Une ln/Une Out 1 the use of alternative models interpretations that reasonably represent the 2 observed data. 8asee &A EYFFeAt kAewleege, seismie seYiaees iA the SGR aiae 3 geAeiaally Felath'ely laiage aFeas, eF seismeteeteAie pl"&l/iAees.

• The 4 i dent if i cation of seismic sources l!ll!fli!ll:!!}:!]!i~n:Ji,ipi,)l,J:!!! in the SCR 1 5 considers hypotheses presently accepted for the occurrence of earthquakes in 6 the SCR (for example, the reactivation of favorably oriented zones of weakness 7 or the local amplification and release of stresses concentrated around a 8 geologic structure).

9 eo 11 Western United States 12 For the active plate margin region, where earthquakes can often be 13 correlated with tectonic structures, those structures should be assessed for 14 their seismic and surface deformation potential. In the Western United 15 States, at least three types of sources exist: (1) faults that are known at 16 the surface, (2) buried (bl iAe) sources l:liillmRI:iil!i::::::11i:mini:lf:i]~:lil!:[ii:!liil:l!i 17 :ititi:11:tiiit.lJ::,t:1:tiil!iileI:1w:l (3) subduction zone sources' such as those that 18 exist in the Pacific Northwest. The nature of surface faults can be eeteia 19 Rt-i-Aee @iliJ::91:iil by convent i ona 1 surface and near-surface investigation -~ 22 23 techniques to ecteiamiAe iiii:~li strike, geometry, sense of displacements, length of rupture, Quaternary history, etc. Buried (bliAe) faults are often accompanied by coseismic surficial 24 deformation such as folding, uplift, or subsidence. The surface expression of 25 &l-4fiEl lffiliili!. faulting can be detected by the mapping of uplifted or down-26 dropped geomorphological features or stratigraphy, survey leveling, and 27 geodetic methods. The nature of the structure at depth can often be 28 eeteiamiAee ivil:Yli~ii by core borings aM ii geophysical techniques. 29 30 Subduction zones are ttii:(Bisffc seismic sources in the Pacific Northwest 31 and ier=l!l:!@i;l:llel Alaska. +he -s:leismic sources associated with subduction 32 zones are the interface between the subducting and overriding lithospheric 6

NUMARCComments March 19, 1993 Line In/Line Out 1 plates, faults within the overriding plates, and intraslab sources in the 2 interior of the downgoing oceanic slab. The characterization of subduction 3 zone seismic sources should include consideration of the geometry of the 4 subducting plate, rupture segmentation of subduction zones, the geometry of 5 historical ruptures, constraints on the up-dip and down-dip extent of rupture, 6 and comparisons with other subduction zones worldwide. 7 8 -~ 11 12 Before pro¥ieiAg specific guieaAce, the followiAg syAopsis of the ee¥el opRleAt of the SSE is preseAtee. 11:iuf.ii]lt::;:11t,:l:,::tt1::1w::::Jj)ff.1nt:rn:1mr:161[::s.s.:1 l@N!!1Ulllii:en::::::11111:,:1:%]llffin1i]i!lrll:m:u111111§::j1:::111:11UJ:1111:1i:atl!IH@t:r~l!i{:1:i:11:1 13 it1:t:1::111111i::i:1:11:n1iI~lt:1:n:Iilit::1111::1,:1:1,1ttY:lli ~ta::1:t!IBililii:lit:::111ti1:itfl:1f§i! 11 14 ,:1:11;1:::1:1111a:tJ it::1hii:lilRl:::I:1111::11:1::11i1:11iliiii:i:tiilt:11;::11i:@1:tl{l1inl!iii11 1 15 liiiliiilOa:1:1:11u:::1,11:a111:::::::1111:tilllilil]ii!IIiinlllil:1iill:IItMi:b1i]illiiiiirfilili!!ii:t:1::,,, 16 1111~:1:1::::::1:lIIll1::::I:11,::::::11i111:11:i:19.:1:1111::11,:1:::i:1:eiiii1I:i:llHIIifl:iliilif!illili]m1::1::::::1n111:1:1 17 i,i,iij/)ii§ii!Ji:ii,iij~j:m;~i:f::: The ee*,el opRleAt _of the SSE m'=:::::Jl!J:!lfj:ii!!i:J!Jil¥ follows two 18 re~Yi ree, paral 1el paths IH1:::::::1,9m~::11~::11:::::::1mi11::~:: 19 -~ 22 23

1m::::::::::Jt]linlr!iilris.:i111.m:i1u11::1:1:11:i1,11::e1:i:21:J::lIIiilli!IJ::11:,:111::i::::::::111 i iPl!l i?:i!l ::::::::,::n111:1:1:1:11:i:1n1::::::11::r111:111:111:::::::i:1i:11eiili:1::::::1 1::*:1:::1::1::::::11,n:1:1::r.tt:::1n1:::::::11:1]111111:t1:1:::::::111:111:s::::::111111:1:::::::11,::::::1:1:111:::::::1::1::::::1:11:11111:

24 llii.liilf.itiii!1::,1]:2:t1ilr2iili]:lil\]i!inIIiliiillil:I:!llllfill:ttilil:!is.i!liilf::::1:11 25 1:UHin;i:11::::::~nt:::::111:::::::1°J.itl~lllF4:r::::::1:j::1,:i:]iHi.iill.6it:11,1.1.,n::::::u:;.:1m:~:::11bi[ifli.'iJ:l.:i 1 26 e~i1:111::1:~::11::::::111:::::::,11:1:int1:::::::1111::11.:i:@1mi::::::::,,:1::,m11:m:1:111:i:::i11e 27 iiiitiii:t:1111:::1::niii:llli.:illiitii:]iiri]!ii:ili:!Ili:t:1:1111:i:r.1:::::11g:::isniriili:r1:11 28 1;:1:1::Jmliis:i:1:111iiiJ]!iiHllillli]lml!ffili:slilBilii!tili:]:111eii:1111:r1i:1:1 29 S!l!§ll~U:=::f!!l:lt!If:1 30 31 32 7

NUMARCC-omments March 19, 1993 Line In/Line Out 1 2 3 swrmrm::o.e.,1:l:oprnan1e$1mu,r:hu1m1:1:rnf.1tta:t:1:o.rra,1:set:£0.tmthet:s.s.er::1:st:ne.1.,,:uwm 4 :Hil:liiinlllitli 5 6 §Mt:!tt:::t!1nlttilliHl:t111m111J:1lttil:llliI§liit:Ut?IPl!IU:11:11:11:rtit:i:11:1:::rnttll! 7 Ill 8 -~ 11 12 IJII ::iftIP:i!1:1:~:n:1::::::!illt:111::B,:r1ibi]:lii:1:tinll!i::liJfflS!il!tti:1:::::::11:1:11;,::111:::::::1::1 11 IPPini:!:i:III QRe patf:I is FefeFFed ta iR FigYFe 1 as EleteFFRinistie ,oinalysis (ElA) aRd eRe 13 patf:I as PFeeaeilistie ARalysis (PA). Tf:le iRitial step iR tf:le pFoeess is to 14 eetaiR tf:le site and Fegien speeifie geelogieal, seisFRolegieal, aRd 15 geopf:lysieal data. BFaRef:ling fFefR tf:le fiFst step to ElA, tf:le seisFRie SOYFees 16 aFOURd tf:le site aFe identified and tf:le deteFfRiRistie seuree eaFtf:lquake (ElSE) 17 fer eaef:I seuFee is deteFfRiRed. GreuRd fROtion is ealeulated usiRg ElSE~ ~Rd tf:le 18 FRetf:lods aeeeptaele to tf:le NRG staff deseFihed in StaRdard Review PlaR (SRP) 19 Seeti eR 2. §. 2, "'!i hFatery Greund Hoti OR." Tf:le eontFel li Rg eaFtf:lquakes feF -~ 22 23 tf:lis patf:I aFe deteFfRiRed as illustrated in Figure 2. Tf:le iRitial step aloRg PA is to eoRdYet aR EleetFie Pewer Researef:I IRstityte (EPRI) eF a LawFenee LiverFRere Natienal LahoFatery (LLNL) seisfRie f:lazard assessfReRt ef tf:le site (EPRI NP 639§ El, Ref. 1, and NUREG/GR §2§0, Ref. 2) feF EasteFR U.S. sites 24 sf:lould ee perforFRed. Tf:le results ef tf:lis assessFRent aFe eeFRpared to tf:le 25 eelleetive assessFRents ef tf:le eurreRtly operatiRg plaRts as deseFieed iR 26 AppeRdix B of tf:lis guide. Tf:le site seisFRie f:lazard assessFReRts are 27 deaggregated as deseFihed in AppeRdix C ef tf:lis guide to oetaiR tf:le 28 eentrolliRg eaFtf:lquakes fer PA. GroYRd FRotioR eased OR tf:le eeRtrelling 29 eartf:lquakes fFefR PA are alse ealeulated as deseFihed iR SRP 2.§.2. Tf:le greuRd 30 fROtioRs frefR tf:le ElA and PA eentrolliRg eartf:lquakes are eeFRpared ta tf:le SSE 31 groYRd FRetioR or are used ta develop tf:le SSE. 32 8

NUMARCComments March 19, 1993 Une ln/Une Out 1 :J@iYttm:::Pi.rftttttlfflllil!i]::1:1ffllt=i1::11:t:11.1J:<<1:i:tJllMJl¢.:i:1.m,1:g1:11J:¥\ljtilf!lffilP.IYi::J:¢11:: 2 lixi\11:lsii:liiifj 3 4 1th1t::,1.:r.tt.r:1t*p:*:,,1hitne1t:er.m1:rd:n1::ttq:mss.em:::entaH::1t:comts.*filn:s:1=v1*r:1eo~l:tc.:11::;: 5 1,1:1111:B.i\f.ll¥iJl.ffi1l)!Htllbli]i=AIJ::mu1v,,,1;1111;J1d.1@1mm111.M:tl:ll¥i."1Mlit=Ji9,ijlil1.tl 6 iinlti.aliillffllrlit:J:!ti):i!Jii@lliiilR![!§iitP!!l.l!lliI!l§]!ll@t[t!:i!ilIIl:llliix:tii::t1I 7 lli!Ilisllni!@Iiiii]i!IMil!\llllii.Sili!lliellnll\!l!illiiitil:1::::1:1,1:J.:1i:i:11;11111:i11:::rt1i 8 111'.u1!:!:::l1;lla.!:l:!Rl':~:::iil~S:f::):::it::~l~::::j;J:l@11l::1j:j:!:e1=::j:j:i.11¥111:i11il:1,:::::::111:j;j -~ 11 12 f:i:::::::::::::i:m:1:1I111:1:,w11::::;:11ln;.11.,:,:,11:1:a1,Ir1H,1t:1111ua::::::;.a.:H111r:u1::t:1:11::::::,j:=~:1s.~::rn:1*:::::£n1

1:1:11 13  ;::::::::::,:::::::::::::::::::::11~iti.li.ii)/i.ll.li¢l)rnHiii=ili:§11in~1:11::::11r1:ttfflffllt)HliifiiJi])i.)liil1lij[:V.1t1:1:::;:11 14 :1:1:IIIll:i:i:ii:

15 16 i!i!I:i:il:iii!i:I:i!Il@li1:J::111::1:1:1111111t::iiniliil!l.iliiiiiilliiiilili1l!fijijllltiliit:i:rilli!!iliili:i!::I 17 1mt:1:1::::::mi:~::11:): 18 19 The level of detail for investigations around the site is governed by -~ 22 23 the Quaternary tectonic regime and the geological complexity of the site and region. Regional investigations such as geological reconnaissances and 1 iterature reviews should be conducted within a radius of ~ jpg km (200 l{l!I miles) of the site to identify seismic sources . Geological, seismological, 24 and geophysical investigations should be carried out within a radius of 40 1:; 25 km (~ :I:?: miles) to identify and characterize the seismic and surface 26 deformation potential of any capable faults tecteRic seyrces aRd the seismic 27 ~eteRtial ef seismegeRic seYrces, or to demonstrate that such structures are 28 not present. Detailed geological, geetechRical, seismological, and 29 geophysical investigations should be conducted within a radius of 8 km (5 30 mi 1es> of the s i te:1.:1.i:i::::iiii1ini,:ili:~: to Etetermi Re iiilii!'=I the potent i a1 for 31 tectonic deformation at or near the ground surface in the site vicinity. 32 Sites that are located such that there are capable fiiiJ=~!~: er seismegeRic 9

NUMARCC,omments March 19, 1993 Line ln/Une Out 1 stryctYres within a radius of 4G Ii km ( 15 mil es) s~eYl a ~a\le ffi~¥::::::r=isimr;i more 2 extensive geologicjJl and seism-k§liii:~::i,j )1: investigations and analyses (similar 1 3 to those within a 8 km (5 mile) radius). The regions of investigations may be 4 asymetrical and larger than specified above in areas near capable tecteAic 5 seyrces liilllll, i@~!ililii~li high sei smicity, or complex geology. 6 7 1.lot.,~:iilf:ig11iil::f,wtn1tiJ.ii.r.11r.,ii.iiw.1$::1:tict1:ti,,1::Jz:t~,-<.it.:ittl*:l1W:t11it.iPi-i.i:i:i:illiil 8 le:l~lfflfi:l:t:lm~ -~ 11 12 Where it is determined that surface deformation need not be taken into account, sufficient data to clearly justify the determination should be pre-sented. Because engineering solutions cannot always be demonstrated for the effects of permanent ground displacement phenomena, it is prudent to avoid a 13 site that tt:l}t! a potential for surface deformation id::J:J]ii,il\,lJ:Ji:Jiiililf:i 14 iiiili!liiI!:!iil!fl!~liniJ!IillniiiiJlit:~:in-15 16 For the site and the area surrounding the site, the lithologic, strati-17 graphic, hydrologic, and structural geologic conditions will need to be eeter 18 flt4ftee ili!rislin~::11~- The investigations should include the eeterlRiAatieA 19 miiiill.iillii of the static and dynamic engineering properties of the materials -~ 22 23 underlying the site and an evaluation of physical evidence concerning the behavior during prior earthquakes of the surficial materials and the substrata underlying the site. The properties needed to eeterlRiAe !'!!~~II the behavior of the underlying material during earthquakes and the characteristics of the 24 underlying material in transmitting earthquake ground motions to the 25 foundations of the plant (such as seismic wave velocities, density, water 26 content, porosity, elastic moduli, and strength) should be eeterlRiAee 27 mii:J:ijril* Geological' seismological' and geophysical investigations are 28 described in Appendix D to this guide and geotechnical investigations are 29 described in Regulatory Guide 1.132, "Site Investigations for Foundations of 30 Nuclear Power Plants." 31 32 10

NUMARCQ>mments March 19, 1993 Line In/Line Out 1 IE)fNTIFICATION ANQ CHARACTERIZATION OF SEISMIC SOURUS 2 3 "Seismic source" is a general term referring to hoth seismogeAie seupees 4 aAd eapahl e teeteAi e seurees 1,e1iln!:P::::::::11li~::::::ygl[1ill::1,~1:11i:11nliiii:&11[:1:11 5 l,i,p.~~~~t:~,::::::~9-p.µ.¥,:i-: A "se i smogeA i e ~~~:~~1['9 source" is a portion of *the earth 6 that is considered to have YAiferm seismieity !I~ fsame D£E m1.1:,,:m1nt:::m11.1,H:IPI:! 7 and frequeA ey of re eurreAee iil:liillil!lU:liiirtt:!!i!Mllil!:1!I!l1.l:~!iiii:ll!ri:iliiI:!§~ 8 ii~:1:x:1:111::m11::::::11r¥+* A seismegeAie i@~:iml:i source would not niis,:11:,~::]:1 cause -~ 11 12 surface displacement. SeismegeAie $.gJ:imH~ sources eover a wide range &f poss i hil it i es from a well-defined tectonic structurel! to simply a large reg*ion of diffuse seismicity (seismotectonic province). A ,;1:1::111:itiiRiiiiIImixIIllili::tti ii "capable teeteAie se1:1ree !iti:J:l"!;l!i!:::1tt:Jj§9, is a tectonic structure that can 13 generate both earthquakes and deformation such as faulting or folding at or 14 near the iiJll.!~::i surface in the present tectonic regime. Appendix A contains 15 definitions of these and other terms used in this regulatory guide. 16 17 Investigations of the site and region around the site are necessary to 18 identify seismic sources and determiAe J!!l!l!!ii!IJ,i!Yil! their potential for 19 generating earthquakes and causing surface deformation. Identification and -~ 22 23 characterization of seismic sources are based on regional and site geological and geophysical data, historical and instrumental seismicity data, the regional stress field, and geologic evidence for prehistoric earthqµakes. The bases for the identification of -too seismic sources should be documented. 24 Appendix D describes investigation procedures that may be used in identifying 25 and defining seismic sources. 26 27 The following is a general list of characteristics to be determiAed 28 e.vil:Uited for a seismic source: 29 30 1@11::111.@ &;j:ource ~ geometry (1 ocat ion and extent, both surface and 31 subsurface). 32 11

NUMARCComments March 19, 1993 Une ln/Une Out 1 2. Description of Quaternary (last 2 million years) displacements (sense of 2 slip on the fault, fault length and width, age of displacements, esti-3 mated displacement per event, estimated magnitude per offset, rupture 4 length and area, and displacement history er Y~lift rates ef seismegeAie 5 .f&Ms). 6 7 3. Historical and instrumental seismicity associated with each source. 8 -~ 11 12 4. 5. Evidence of paleoseismicity. l1nn:11:1.i:im:i:a11i1::1:is:l::Iii~:g,1:sIIiimm1l:1:lf:tiill![~e1at i on sh i P of the f au 1t to other potential seismic sources in the region. 13 14 6. OetermiAistie Seyree Earth~Yake. (Details fer the eetermiAatieA ef the 15 OSEs are ~reYieee iA sectioA 2.) liiii:ffiiq[!::111!di!M!@l1liiiils1.iiii:!Iili:l![iin 16 ll[ii1lrililIIlililli:Jill:1m1Jtfiliisi:I: 17 18 7. Recurrence model (frequency of earthquake occurrence versus magnitude). 19

8. Effects of human activities such as withdrawal of fluid from or addition of fluid to the subsurface, extraction of minerals, or the effects of 22 dams or reservoirs.

23 24 9. Volcanism. Volcanic hazard is not addressed in this regulatory guide. 25 It will be considered on a case§byt:case basis in regions where this 26 hazard exists. 27 28 10. Other factors that can contribute to the characterization of seismic 29 sources such as strike and dip of tectonic structures, orientations of 30 regional and tectonic stresses, fault segmentation (both along strike 31 and down-dip). 32 12

NUMARCC.ornments March 19, 1993 Une ln/Une Out 1 11~i:i:i::1:1:i@:~:i:~:g:~i@l,::i::::i:ri::i::::~~~i:i:::~11:~:i::::::111::::::"1":1m!:1::::::'=!111~~:::::i:1~1::::::1111::::::11,1::::::1,~1::::::§@j:~ 2 1eei111a:::::::11:r:1ffi']:sr;1:1,§t::iij1;um::::::,:11.::::::t.vI::tiitJ.ma.1,::::::111.:11:1:11:::r1.111r1:;:r::]o,i:::::1r:,:l:1.e.e 3 ~:e,e.!:fi~:§.}~!~)~j:~:f~li~~~:::=::~~:~:~~:~;f~p.µr.¢.@:,WAfi'-@1~~!:r-:tui~~~..~~i;~:~:!-~::::i:~?:!~~~i~p;~~l?:'-9 4 lin:11::elllimiIIiiii:1::,1:,nei]liil!HIIniiililfiliIIllif!l)iilil!i!Iiii!llbiIIiiilllinilIIiil:ti:i1:1 5 :1liiliill1:11t:lilli 6 7 2. BfTfRMINISTIC SOURCE f.OiRTMQYAKfS {()SES) 8 9 gs Es are the ~J::::::::~1t1:::::::1w::::::111::I11:E~mi:s:::::1:iiins@Iliilili~lin1:1i1i:i:1,:i::rn::111:::::ffii~1:~1 -0 11 miiiill!Miil~:1:1::tli!iliitirm:liii!lllI]tni::::::tiiilllmlt/llJii§n:1:l!tliit:1::iIIlni 1arge st earthquake~ that can reasonably ~e eMpected te occur in a given seismic 12 s OU rcelilllsiniiiiliilinlfilli!§llflni:i:;:11:cjiitliilin:1:e:::::::,11r:iiilleiJisillbiriillnii:iiliiiiilili1I 13 -ffi the current tectonic regime. Betermi Ai st i c seurce Jli1ffiiii:ll);:j:ffiil!ii!llH~~ 14 earthquakes are characterized hy their magAitudes aAd, as a minimum, will be 15 the largest historical earthquake associated with each source. A larger 16 earthquake is warranted when specific geological evidence is available, e.g., 17 paleoliquefaction evidence of larger prehistoric earthquakes! or when the rate 18 of occurrence of earthquakes indicates the 1i kel i heed p~~jpJt:j;:1,J: of a larger 19 earthquake than the largest historical event. -~ 22 23 Stable Continental Region (Eastern United States) In the SCR there is a short record of the historical seismicity and 24 considerable uncertainty about the underlying causes of earthquakes. Because 25 of this uncertainty, it is necessary to use ceAsiderahle judgment and a 26 variety of approaches to es tab 1i sh the 9Sk ffia.x:Unffiii.Jiii.grM/)~ua~t:11r:t1::ts/e.jfjiml)e 27 lf.;P!!S@* In addition to the maximum historical earthquake, the determination 28 of the BSf earthquake mi:i:,:11m:::::miln:i1:1i for each i dent 1fi ed sei smegeA; c 29 1ifi:J'iiii:c source is based on the pattern and rate of seismic activity, the 30 Quaternary (2 million years and younger) develep~eAt aAd characteristics of 31 the source, the current stress regime and how it aligns with the known 32 tectonic structures in the source, and paleoseismic data. Bl!iffliiiIII 13

NUMARC Comments March 19, 1993 Une ln/Une Out 1 RtiMllil.@fl:::1:U:IWUtt'.l}ifittm1.11:1rm1i(JJjJ!t:111m1u1::::1,ttllUlll.i!%Ui:111:111~= 2 3 ffli.t:e.:11:tatt.$.§P,f::t1,1.,:1:mwt;:;jJgQJil.Mle.t:iot.Mi*itt.P.t.i1=:::=[111:1.i1:¢::tlAQt¢.i,$J]::¢.lt\t:lffl 4 ,,11.:mt:n::r1D1aill§,mi=s11.11inlil==e111,11:;r::=11Rcs11i1,111fi11~,n11:r.*1Prxt~**:i@J.1u,r.!1 5 1111,1:11u:J:1:::::1,n111:i1:=::::1::~: 1 6 7 Western United States 8 -~ 11 12 In the Western United States, earthquakes can often be associated with tectonic structures. For iii.iilli faults, the lixl~ffi magnitude ~ earthquake is related to the characteristics of the estimated rupture, such as the length or the amount of fault displacement. The following empirical 13 correlations can be used to estimate GSE-s illlEli!lm:::::N111:1:1111 from fault 14 behavioral data and also to predict the amount of displacement that might be 15 expected for a given magnitude. 16 17 Surface rupture length versus magnitude (Refs. 3-i Iii>* 18 19 2. Subsurface rupture 1ength versus magnitude (Ref. + l:Q) . -~ 22 23 3. 4. Rupture area versus magnitude (Ref. 8 :Ill). Maximum and average displacement versus magnitude (Ref. 9 :l,li>

• 24 25 In the Pacific Northwest and Alaska, GS& lfiilI:ffiillllli:Iffiilimt1tl must be 26 assessed for subduction zone seismic sources. Worldwide observations indicate 27 that the largest earthquakes are associated with the plate interface, although 28 intraslab earthquakes (e.g., the 1949 Puget Sound earthquake) can also be 29 1arge. GSE-s JlllRiil!mfRsl1.:1,11 for subduction zone sources can be based on 30 estimates of the expected dimensions of rupture or analogies to other 31 subduction zones worldwide.

32 14

NUMARC Comments March 19, 1993 Une ln/Une Out 1 3. PROBABILISTIC SEISMIC HAZARD ANALYSIS 2 3 1:~:::::;[lilffalillilU:illl;:t:,:111:;.;:tuiii!II!lii~i\iliii:ll:11111: 4 5 A ,rohahilistic seis111ic hazard aRalysis (PSHAt should be carried out for 6 the site. A PSHA allows the use of multi-valued models to estimate the like-7 lihood of earthquake ground motions occurring at a site. The PSHA systemati-8 cally takes into account uncertainties that exist in various parameters (such -~ 11 12 as seismic sources, maximum earthquakes, and ground motion attenuation). Alternative hypotheses are considered in a quantitatjve fashion. The PSHA can be used to deter111ine iil:J.!ll.111 the effects of varying significant parameters,

1,1 to i dent i fY sign 1f i cant &ii~ii:t11:1:11::::::~1:11,11:1t:]sl::::::i1~::iffii@ sources , 1n 13 terms of llitl,6§'.Y.li.ki magnitude and distance, and to previde hazard esti111ates 14 for Yse in seis111ic prehahilistic risk assess111ents.

15 16 The results of a PSHA are specifically used to lil@im~:iiilli!iIIlll!IIt::1111 17 :11mmm11t::::1,11ii:1,::::::1:;: : 1:1:1=:1miilIIl1ilrtiliii;f9ili~l;iiin1!M1!1::::::1111;:;Iliin!ili1:11::::::;~1, 18 iiil!nl:111:1:11I::11i:111i:111:~:::11:::::::1:1'=:1:i:!ID.illi::li:111:a1:1::::::inriiii:i:i:iilili!iiltiiiiliil:Iirn 19 iiri~!ni.!i:Ii!ii:~::1!li::::::1,,11:1::~li!llinii::::::~miml!l§i!:: : :!:l:11:~::::::::::::::de r i 'le cent re 11 i ng earth~ Yake s -~ 22 23 as discyssed helew and in Appendix C. It can alse he Ysed to esti111ate the annYal prehahility of exceeding the SSE aRd de111eRstrate that the annYal ,re hahility ef exceeding the SSE design greynd 111etien at the site ee111pares fa'lerahly with that fer the eYrreRtly eperating RYelear power plants. (The 24 preeedYre for this de1110Rstratien is deserihed in Appendix 8.) 25 26 !tlt!!l:~11l!:tlfl~f:MidinliU!!Pllll!!@:n1!1.!4.I'll!]'ll::::::§::l~![i1:n:::i:!l!l:::tlPH ei the r the 27 La"reRee Li¥er111ere NatieRal laheratery (LLNL) (Ref. 2) er Electric Power 28 Research Instihte (EPRI) (Ref. 1) seis1Rie hazard analyses, j,9]li1iii:eli~::Illl1 29 m,~111,1:111:::::::~::1sm~1:1: n1:J1111::~m~:e'l!9!9!11~):::::J.fil]!l:!:t!mll!t51M as see i ated data has es , 30 should be used fer plant sites in the SGR. However, alternative seismic 31 hazard analyses may be used with prepeF l!l!i.M!II justification. Fer the P&MA, 32 Yse of the seismic se~rces identified in the LLNL aRd fPRI stY~ies is 15

NUMARCO:>mments March 19, 1993 Une ln/Une Out 1 eensidered aeeeptable eHeept in regiens ef the SGR *11th high aeti¥ity rates, 2 e.g., near New Madrid and Gharlesten. In these eases, it is neeessary te 3 either deseribe additienal site speeifie seismie seYrees er she*1 that the 4 regienal seismie seYrees in the LLNL and EPRI prebabilistie stYdies adeqYately 5 111edel represent the teetenies in the ¥ieinity ef the site. :Illlii:[iiffii[:1:1:til 6 :11:;.:1@M:t111rm:pfi.1:tNl!fffll@:~~:11::~gj!)!!!~)jP,]j)~l!!m~IIf!!~!Plllt:tP.P;:qr1:rm1@@l:1t:mg:111 7 liil!l:ri:lm 8 -~ 11 12 lifiilill1liill:::::P:Probabi 1ist ic methedel egi es ill:!ffllllilllliiiflll!~iffi!liliil~l simi 1ar to -the 1:nPi~:::::::11 LLNL and EPRI sei Sllli e huard stHdi es::i:: have not been performed for the Western United States. mliriliii:tl]:i~or Western U.S. sites, a site-specific PSHA must be performed and documented in such detai-1 that a thorough 13 review can be carried out by the NRC staff {Refs. ~ lJl:IIIIJ[{lll). 14 1s 1:r:[::::::::::::1:::111,r1u::ne.:[1n1:1:1,:11:1ss:m 16 17 111::::::111;:::::::i:=i::::::11111m1::111:::::::111,::::::1111::::::1:1:11::::::11:,:1v:::::::111:111::::::11,11::::::1:~:::::::1::::::1n111111:1::i111 18 m,11:1:t:IPiiJ:i:1111:Jw:1:1.:1:Jil1JRl1:in:r:111:irll]Pi!:::::,11::11::u~1:::::::,i:1ni!if:f:fiilI:Il1ii}IPlin 19 l:~:iiliilf:inl1!:IUllI:::1:1:1::tt:11:i::i:I:liiiiiH~IIl:i:lifleii~:i:;::::::1::::::igt:11:1::1:::Ilil.Hll:ilii.it::111:11 -~ 22 23 liili!ll!i:i:Iii:l!I:iiililiisiI:::1!!1:11,:~:int:!tirf:iilll:iliif:ll@:llilininsi!::::::11,111!1:J::1:11::::::1!

   ~~e.~~:J:p.9.::::::~,~t:~~~@::::=::::::

4. GONTROLLING EARTHQUAKES 24 2s I* 1i11:i:,,:::::::1::::::11i::1mi:1::::::1:1:1111::::::::~:1:itrlitllllnI:::ililiillliiI:111::::::11:1 26 27 IIM::::::PMEPl:~@:j:j:jgt-)j:j:j:~11:t:!!ll:!ii:~:s::::j::§11:1rlt:j~:jqf,gffli.l:~:jpg:j:j:jl,1j1,j:j:j:jj,j\j:j:j:j!,gjjpr,gy:Jjl1::i:i:j 2a 1.wn,r1::::::put1:o.1.::::::1:1z,11::::::r.e$.:P.lt:~@::i,i::jj::1,tm,:1:i1.1:,1:t11,:lm1::111j,j:::::=4.n.:t,ua'-r1:,1:,n,:1::nst:01 29 ~b(f{ll:J::lfflJ:~f::l:mnmJ~§H::~nl:::::11m.:n.qggk1.$l::~H~n;::::::=ire.n::lbtt!§a:$.:l:#,filttm::t.b~Mf$.llitml:s::::::t:~1,:J:9:J:i 30 1,riqijmb:*t't:1n;~::::m~:::::11p1na:1:x::::::1::::::1<<seH1::he.:,:::Jt.ttet::1r°'1.a.u:r,::::::10.r:::=::1e.ve.1:,,:1:ns.=:::::1:::::::s.1.:1:1m1:1 31 h1tt)i:l::r1:nfitrffll.t:1:~n:tll.$tht:1na@tn~/fl$.P.~¢i::f1lt':tres.:li.l:f{$.:l:tha.tr=:$.b.~n.tld.t:b.e.t=1.ro.l:il.tldt:'fiY 32 ,:m~:::::::n,ii.ra.:::::jj'ijili\$.l:il$]i 16

NUMAAC C,omments March 19, 1993 Une ln/Une Out 1 11,::rn1.~.u::~1n:i.1:1i.z.1m::::::1:1:1am11:~1@:1,,1.:::::::,::1::::::1,,1.1:,ua.::::::11.]=11,::::::111::::::1n~::::a:11::1.1:,u 2 ~~::::::~:~~::::::1.~ii:f:i.g~p:~:9-ij::::::i.~::::1111:;;lli.~!fl:::::::1:~::::::11*::::::pE:l:l:l~~~t.Y.!~l::1::mi.in:~:'-:"-~]::::::,,~ 3 d.j:Jfiriii}IiHttiiil:UiliriiifiJ:iiiitiiii:i:i:hmU:ii:iini:l::ij:lil¥ 4 5 ,::i:::::r:i::t:::111:111It!t:11111:i111:1:1r1i11lllllHEl!:!l:1:111rll11::l1t!lill]Jtll!illb!1 6 PIPPiPli!l!l}!ll1!]i!!I!iiiiill:n1:Imlit!llll!i:! 7 8 -~ 11 12

t::::1:::ttflliil!il!lii:Iliiilt]:ll!lli!i:n[:§!l;::li11:n::::::,1:1::1mi!e!:Ii11r1i=I:ll§!]!l:l1:i:Isilli,::i:11

1:111:11:::i::::::1111111:

13 GeAtrelliAg earthquakes are these earthquakes that ha¥e the greatest 14 effeet eA the greuAd 11et i eA at the Auel ea, pewer pl aAt site. -~~){ffl!i~il~~~~ 15 l.!~J:11)!£11.S!!l)!Ef!lnt:19.Y!i~!!)J)!)9!!)1:l::::::§in:11l!!!,,::m~l@IIR!lt§PJ.il::::::b,:j:9§::=:J!i!)tl:a 16 liililiix::::::miiirit:1::::::el']i.minl::::::m11:,::enl!f:! The Fe !Ra:)' ee se*veral eeAtrel 1i Ag 17 earthquakes fer a site, e.g., a l,fRoderate nearby earthquake may control the 18 high-frequency portion of the ground motion spectrum, and a large distant 19 earthquake may control the low-frequency portion of the spectrum. See Figure -~ 22 23

   ~       lii1t1J:1:r1IIli1snl!§1li]!llliitinmieiiil!::l!li:tlil!imi:lii!t:l!ni::l!::li:ll:!1mi!i:::Inii.lil ll\l!tl!l:l]!i!lll!ii!Iiinl:J:Jlb!ttmlttifJ!illl!Ufiffill9!1:IP.lfffll!i91!il'il:111=1=1m:J11::1:111,:1 IA the Beter11iAistie AAal:)'sis (Figu,e 1), the eeAtrelliAg earthquakes 24 are deter11i Aed h:Y the fell e~1i Ag preeedure.

25 26 Fer eaeh seis11ie seu,ee, plaee the BSE at the elesest appreaeh ef that 27 seuree te the site. Fer the seis11ie seu,ee iA whieh the site is 28 leeated, the BSE sheuld he eeAside,ed te eeeur at aeeut IS k11(911i) 29 fre11 the site. 30 31 2. Beter11iAe the BSEs that preduee the largest g,euAd MetieAs at the site. 32 GreuAd 11etieAs at the site frem BSEs are esti11ated usiAg the preeedures 17

NUMARC Comments March 19, 1993 Une ln/Une Out 1 EiescribeEi iA StaAEiarEi Review PlaA SectieA 2.5.2, "Vibratery GreYAEi 2 MetieA." The earthqYakes preEiYciAg the largest greYAEi metieAs at the 3 site are the ceAtrelliAg earthqYakes. 4 5 IA the Prehahilistic ~Aalysis (see FigYre 1), the ceAtrelliAg 6 7 8 Perferm a prehahilistic seismic hazarEi aAalysis fer the site. The -~ 11 12 2. aAal ys is wi 11 Eievel ep YA i ferm hazarEi spectra at se*.*eral aAAYal prehahilities ef eHceeEiaAce. Qeaggregate the prehahilistic seismic hazarEi aAalysis resYlts te 13 iEieRtify ceAtrelliRg earthqYakes; their EiescriptieA iAclYEies magAitYEie 14 (M) aREi EiistaRce (Q) frem the site (see AppeREiiH C). This EieaggregatieA 15 sheYlEi he EieAe at the aRRYal prebahility ef eHceeEiaRce level EiiscYsseEi 16 iA AppeAC:ti>< 8. 17 18 The ceAtrelliRg earthqYakes thYs EieriveEi frem the EietermiRistic aAEi 19 prebahilistic aAalyses caR be cempareEi at this stage te EietermiRe whether the -~ 22 23 ceAtrelliAg earthqYakes frem these twe appreaches are similar aREi alse te EietermiRe if the ceRtrelliRg earthqYake er earthqYakes that will EiemiRate the greYREi metieR estimates at the site are easily iEieRtifiahle. If the EiomiRaRt coRtrolliRg earthqyake caR he iEieRtifieEi, the groYREi motioRs are EietermineEi 24 eRly fer this iEieRtifieEi ceRtrelliRg earthqYake. If the coRtrelliAg earth 25 qYakes frem the twe appreathes are Eiissimilar, greYREi metieR estimates are 26 maEie fer varieYs coRtrelliRg earthqYakes anEi cempareEi te Eierive the fiRal 27 greYREi metieR estimates for yse iR estahlishiRg the SSE greYREi motieR or 28 cempariRg it with the SSE greYREi motieR. 29 30 1:~f::I::::::1::::::11111!lfil!!1!li+/-ll@P.!!1111!i::I:liiI:::11:::::1P:eiilil::::::1,:i:1mlle:11,uursi:t 31 32 18

NUMARCComments March 19, 1993 Une ln/Une Out 1 mu11::t11:1:1t@tll:sit:h'~)M$))UMii,:J~hi;tH]it1:111;1p.tabl:l(W[,Jf~;$det1tm:UU:pgt=at:S.S:&Wl 2 loof:t:nnatao.rr:i:p.:Jt:;tn1:sn:pne}n~l:k¢¢.e1-t.1n,il::::11M:1ms1*1:1::JJ;c:::::::,1:tgw,J.:1.n:obt1;ftned 3 by:+coilp¢.t.'tng::::::,nt:,ei'l:tlfuat.ifoi11!i:UtAtiliis.ir,:ateit:&fi9tmdat.1kol~i:1:ni.d::=rlrom 4 tle:tslffie;ti=hve.=s.:t1sa\i:o.tt::dl;ae.~:gt{o.t1tptdii1ltt:lb1Jfo.mJ£J:n#t11iatHif:o.u:ld 5 n1t'ei1=1,ttat.s:it&v,=,:1,o.n.::::,:1rt,c,1,w,1.1,,1a:ar,11:1mI::,t:stittr:1,s:t,1m:,:,::J.~e1:,r 6 tB,ra<**ttz.a.t.,1:o.n$.1;wnde.n11xmed1esgt111.t,:~,a,111:na~pe;1:1t:<<@1.v1l:uat:J.:Ji.n:rt.tt1t 7 ian=tflt:e.1n1,m11rn:t;.;:::r11*tlnn,w1.t1n:1:1Jfr1irM1.11::1.1:11,1.m:,11:,m.i:¢t:$,-ur¢1:$@ 8 11 12 ms.,::::::1.s.a:t1.r1um.t::::r*1.nAn$.ij:::,:::,i11e.1.runit1:itia~t~:rm1:nii::,J,.v:i::::($.c~d:::1:ns::::::1.::j$.:,:1g:f:,1et:.:1::1:1:i. 1

t<<llffihill:l.ffl~\i.Wfflf,11\lii.\l@tp.til:if::$.t:1tt~i'.t4.::::::rt,i.Ait'111\\f1P@Gl'tU.lfi]'ih#l~l:lit:fi.l~ftlS.:I 13 ll!lllillBI:'11,)!!f§IIJM::5:~11tllilPIJ:!1i::II::1,1::;::11:1:lll:lfllel::111r~:11;M'1:lllll~IIH!:§1:~fl:I!

14 -¢e.-1-tau1:1.m:::1roceu.u.tf.::::::t~1:11.t,m1:n~trt1~:1ss1=;1::giwuo1tmr1.:$.1,1n$~lt,1:¢t1Wtiii 15 16 C. REGULATORY POSITION 17 18 During the site selection phase, the preferred sites are those *1ith a 19 IR 1R1IRYIR 11ke 11he eEl e f ;f:gr,j:=1~:ti1.;~l:;::~b.1r:!)::::::}:1;::::1jj:jg11l::i!s.i1:il::@::::::s11i1l:~i:J.::~::!1;;:::~f:! surface er Rear surface deformation or the occurrence of earthquakes on faults in the site ¥iciRity i.!~~ (within a radius of 8 km (5 miles)). 22 Because of the uncertainties and difficulties in mitigating the effects 23 of permanent ground displacement phenomena such as surface faulting or 24 folding, fault creep, subsidence or collapse, the NRC staff considers it 25 prudent to select an alternat ive site when the potential for permanent 26 ground displacement \J:l):Jfgjji\{11 exist, at a site. 27 28 2. Regional investigations such as geelegical receRRaissaRces aREi 29 1 iterature reviews let):::::j!l;lil1l!fiii::it,lm~;il!1iliiiii: should be conducted 30 within a radius of~ I.II km (' J[l'.i mil~s) of the site te iEieRtify 31 seislRiC S8YFCCS. 32 19

NUMARC c:omments March 19, 1993 Une ln/Une Out 1 3. mffliii!ii1:illifiP:ilI)i~!fl'; GQeological, seismological, and geophysical 2 investigationj; should _be carried out within a radius of~ Ii km (~ !1:5 3 miles) to identify and characterize the seismic potential of capable 4 teetoAi e aAEI sei smogeAi e sourees 1111:1:~m or demonstrate that such 5 structures are not present. 6 7 4. Detailed iilil~Miit!! geological, geotechnical, seismological, and .~8 11 geophysical investigations should be conducted within a radius of 8 km (5 miles) of the site to determine the potential for tectonic deformation at or near the ground surface in the site vicinity. Geological, seismological, and geophysical investigations are described 12 in Appendix D to this guide, and geotechnical investigations are 13 described in Regulatory Guide 1.132, "Site Investigations for 14 Foundations of Nuclear Power Plants." 15 16 5. Sites that are located such that there are capable or seismogeAie faults 17 within a radius of 4G ~§ km (~ ;~p miles) shoulEI have ffiij;;;;;i;r,;~99:;J!:~~ more 18 extensive geologic and seismic investigations and analyses (similar to 19 those within an 8-km (5-mile) radius) IIJ:1:1,11:r111t:iti1- The area of investigation may be asymmetrical and extend beyond 4G II km (~ 11,;i, mil es). 22 23 6. Seismie souFees that weFe Aot coAsiEleFeEI iA the LLNL OF EPRI PSHA shoulEI 24 be iEleAtifieEI aAEI chaFacteFizeEI usiAg the iAformatioA ElevelopeEI by the 25 iAvestigatioAs. AlteFAative seismic souFces shoulEI be ElevelopeEI to 26 iAeorporate a raAge of iAterpretatioAs, aAEI the bases for the 27 iEleAtificatioA ef these sourees shoulEI be ElocumeAteEI. Souree zoAe 28 geometry shoulEI be ElefiAeEI foF eaeh seismie souree. For faults, the 29 type of slip, leAgth of rupture, amouAt of ElisplacemeAt peF maximum 30 eveAt, aAEI area of the rupture suFface shoulEI be EleteFmiAeEI. 31 32 7. 9etermi Ai st i c souFce earthquakes, ,1h i ch aFe the best j uElgmeAt of the 20

NUMARCComments March 19, 1993 Une ln/Une Out 1 maximYm earthqYake that eaA reaseAahly he expeeted te eeeYr in a gi¥en 2 seismie seYree, sheYld he defiAed fer eaeh seismie seYree. 3 4 86,. A PSHA for the site should be performed to estimate the annual 5 probability of exceeding the SSE. 11:1::::::1:1:111:::n:111111::];,:n::::::111::::::1¥:B:~: Either 6 the LLNL er EPRI prehahilistie iisii~R seismic ha2ard analyses i@iiili 7 w-4-tft ii associated data bases sheyld §ii be used fer plaAts in the 8 Eastern UA 1ted States. 11::::::1ssiilili[iliilmi:,i:111111]:1:~11.1:1::::::§gi]i~l!1li!I 9 ltt::Ui:l)i:mu@J.:fi.~Utl:J::tt1m:,:p,q:1::fii:A?Q@~O::qg:}.:Sll@:::1::'=1:1::,mqJ:a:1:111l&::rn:1n1:rn1,,1.:tty,:i:1*J: ea 11m1:11~urng~:!:1m1::mim:m11:t::m:1Hr¥:r1nttY:'-:~sm:::1rm11:iAn1§~::1::::Jt~1!:!:1:r::11r:*ati:Un111rn::~,~ 11 1:1::t:11::1:1i1i1:;:::::::::::::111t:1:1:1::::::1::tlll11H::11:1IIii[iiiiiliilIIlil:iv1:J.::1ililIIIIIIiiiiilil 12 111::1mi:sH:1R!t!1:1::~::'::i:j,j:i:::1:~i1:011!:i121:r@1:Imn11:remft111]:gn1:::::::111:::::::ne'=::::::e1ii1:1::j,11n1 13 11:11::::::1:1,:::::::1m111:1:,11:1]ti1;:::::111:11:::::::11u:1:1::,1;1::::::::111m:;,:1:1:1'i:11:(:1:1::::::::1:11:1::11:it11 14 jj11,jj:jjgijjJ'i[fjij:ji:~m:1:t:fijj,1,:tP:!i:j1,:jpjij:t1t1.:lll!1ii11t:mi1if~jj'i[j:jr;j9.ij:1:J1!~:!:I@!)flf::1j1::t1 15 1111:1::11::s::::::1:ii:1::,J::t::I: 16 17 FOr i Ji i i:::::::j::1::::::1:~:, Western pl ants g::*:~:f::::~::~:::::1,:11f:1ii i :~:1,:1::::::111::1m,:1:::::::~g1lil 18 i~iPN:igirt!llif~illili:fifiil:!19. a s i te-speci fi c PSHA sheYl d fliii be performed. Tbe. 19 :1:1111:1::1J:ii!Hii:l:Iinl:t:211:riilitU:1:1.1J:gr[tirol:11:1::1m:1:1::::::i111si:i:::I1ilil:1:::::::11 20 11111;1::1:1111:]11:,:,it::11i:::::::,:,11rm11:,:11::::::1i,si:1:::::::in::::::111:11i11:1!u::e,mt 91 111::1m1::"1:i:1:11::It::111::::::1111111:1:l~11::::::::1:11,11:;,:1,1:1:1n1@ 11:1,m:11:1:1,::::::11:i::1:mJ:1: 22 1.uu.rt.~ifi:t111u.1:d.t::ti1::::::u.e.ve.:l]U~edt:td:twne.nri&nate.:t1r:11n11::r1ntmnte:t11e.mat:uHt$.:;: 23  !!!lttl1II!1:1:1:1:r:11nilb.il[j:9ijij!it:tt!IJJ)pg[Jf![:Jliiii[:1:e1rfiiiI::i!i!9Y)lii[JI 24 lii!ilnlil:IfHIIIHl::ll::ll:tixil:i:,:11::::::;,m.1:r:ineir.eeni:11::::::11::11:mnill!ixlilill!iioliilriil 25 :1:1:111e1,11:11:11:11:1wi1.111::11:i:s1mi::::::::11:i:1m11:11,:i11mi:1::1n1::::::1111:111J:s11: 26 l:iilii!il:~j~ti~fjji:jpjjljijijipjfi:lfii:i:iipiiil:Piiii:~:i:11::::::ii:iiim!:P::fiiiitiiiijijii:il~iisihUil 27 :1:1:::::::11X:iiinsii:I::1:1i::1J1::111::::::1:I liiiiiltiiliiliitlilP:1J:1::::::11::::::11si:nilI::llinfii1I 28 11i:11is::::::111re1::r:::::::::::::11r::r11,1:1:1Mr111::::::1111::::::1tt::1m:i:1:i::::u::,n111::::::1rm:11111r~fi::::::::1m11n1 29 ilitl:i:1:11:11111:1Iiiin:I:Iiiii:!lffl!I]~iinlMIIi!IIIiirii::::::11::::::1111111iiiiiiI1i:rtriii 30 11.iil!II::iii]:liliildiiilif::::::: 31 32 Use the PSHA te identify seYrees in terms ef magnitYde and distanee that 21

NUMARC Comments March 19, 1993 Line In/Line Out 1 eetttri ~yte s i gA i fi eaFttl y te the se i smi e hazard at the site. ffi~l!ime.1~~ 2 IDillllili]!iiiltiB.:t::::::11:::::liiii!iliii!iiil!ii!IIlliiil!j,illilitI!I!Ri:I:irl!ijii:i:Iint:t:11::p 3 ,11:,1:1.1::ttlii.i.I;riiil'J::i;iil.1::::::111:::::11,v,1,1:9.ii.1.t:11::::::,r:::iS.ilimiiJ.lii:iii.lU:::1:1:r,mii:1:1n 4 1,1,::::::1na:t1rov:1:1,:::::::1:1,:,:,1n;::::::1.a::::::t;1,::::::1,~::11ie.:i11u,,.,11:I11~u:]ia.r1Hai.11,i::::::11,1 1 5 lillli:ll!llwliflfflIIlliit 6 7 Z:f[t[i[llll::JIIIIJ::~::1:::::1111irtn:111::::::11:11:,:1:111:1fii::Jiiili:i1Iilliiiti!:§il::I:1:1::::::11ii:IJ:1::::::1::i::t

1:1::11:1::::::111111!:11u:::11:::::::1::::::111,n,ilit:&1111::1::1:11::::::1111:::::::1::1::::::1,ttiiiillinll

.0 8 9 l!lilltM::11ttl!lf:~=11,ttl:l!IIfll!!!llJ!J:~]ll!ttJ:11t=['i!IIIlilt!i!l:J::11::r,:1:1::11:i1:::n1111m l!iiiil:!II!lir]iliiiilti!ni![il:!inl:i::::::=11::::::11:[t.:111111:i:xt:[liliI!il!:IltlliI:ill:1:111::::::::1!llilllf: 11 12 91

• inl:t11:i:1m!:PlllaiilII:in:iirtm:1:,:11t::11:11IIiirIIIIIJ]i;1:1:1::::I!::1[:ililil:iiiiitii 13 liit.s:1g11:l:a1:::::::111::::::111111:im111::11:::::::1:119p;]li!!!t&itll!ii:i!J!iinIIIJ:ilIII:tlimnlii:1t:1:::::::19:

14 11:1::il:iii!j,'iilli;ili:!Jlil:I:l:ii:liitMll:li:i:Ii:nllif:il.ii:l;II:::B:I::iriiii:lll:III::tu::1111 1 15 )jjq)fqmJ\lji;gp;Jl,;iJ.;1\i,. QeteFIRi Fte the GEs that pFeeh:tee the l aFgest gFeYFtd 16 metieFts at ttte site. GFeYFtd metieFts at ttte site fFem GEs aFe estimated 17 YSiAg ttte pFeeedYFes deseFibed iR SectieA 4 ef ttte QiscyssieFt sectieA ef 18 ttti s gYi de aFtd StaFtdaFd Revi e\<i' Pl aFt Seeti eFt 2. 5. 2, "'Ii bFateF;Y GreYFtd 19 MetieA." -~ 22 23 IM=]J]t]Mt@li.ili.it:t:hillS.S.E~tir.i.ilbtfJ=fis.ijods.e::ti:p.ee.liU.liitt}tliillS.S.E!tg=ro.U.hdt:re:i:ijb.h:1-i 1t.l~i!P.li¢,liiiR:illi!j1m:tn1@:J:J::ijj:!:IItl!])ji11:!=!I1::::!:1!t:!11i:1111:t:Jt11!,t!r11P:11:~1::::::11iP:IIP:ffl 1l:1.i1:i:11i:l1iir1::§1it:::~:1::::::11~iill~ili:Ii1tt11:i:Jiiis.:µJ1:1@l:~IIiitt1t:1it~l:1i~j:it11iPiU~1@ 24 llill!~t:1:lii!ille:tt:llite!IIIll+l!tli:lllt:t!:tt::::Rtll!:1:11r.ltI!1:1J:tl:t:1:1::::::z:1::::::: 25 26 D. IMPLEMENTATION 27 28 The purpose of this section is to provide guidance to applicants and 29 licensees regarding the NRC staff's plans for using this regulatory guide. 30 31 This draft guide has been released to encourage public participation in 32 its development. Except in those cases in which the applicant proposes an 22

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 acceptable alternative method for complying with the specified portions of the 2 Conunission's regulations, the method to be described in the active guide 3 reflecting public comments will be used in the evaluation of applications for 4 construction permits, operating licenses, early site permits, or combined 5 licenses submitted after the implementat~on date to be specified in the active 6 guide. This guide would not be used in the evaluation of an application for 7 an operating license submitted after the implementation date to be specified 8 in the active guide if the construction permit was issued prior to that date. 23

NUMARCQ>mments March 19, 1993 Line In/Line .Out 1 REFERENCES 2 3 ~ 1. Pacific Gas and Electric Company, "Final Report of the Diablo Canyon 4 Long Term Seismic Program; Diablo Canyon Power Plant," NRC Docket Nos. 5 50-275 and 50-323, 1988.* 6 7 -11* USNRC, "Safety Evaluation Report Related to the Operation of Diablo 8 Canyon Nuclear Power Plant, Units 1 and 2," NUREG-0675, Supplement -~ 11 12 1* No. 34, June 1991. Electric Power Research Institute, "Probabilistic Seismic Hazard Evalua-tions at Nuclear Power Plant Sites in the Central and Eastern United 13 States: Reselutien ef the ChaFlesten EaFthquake Issue," NP 639§ Q, 14 -°it)F i 1 19s9 111:::,:::::;~:I:1:~r1i11:::::::11:~::::::~1:1:I11:1§ti1:~1::::1111::f:: 15 16 ii. D. L. Bernreuter et al., "Seismic Hazard Characterization of 69 Nuclear 17 Plant Sites East of the Rocky Mountains," NUREG/CR-5250, January 1989 . 18 19 1:i::::r:::ii:::11:1si11:1:11111IIliliiilitill:i:1.:1:1:i1:s11:1::::::::r111:1P:ll:i:Itin:11,:1liiliillli.ilili1~imw.1 i1illa11111IIi:lil!li:::::1,n1:n1:1::::::::111J:11111mrM:itli:i:1J: '22 23 24

   ~ .           D. B. Slemmons, "Faults and Earthquake Magnitude," U.S. Army Corps of Engineers, Waterways Experiment Station, Misc . Papers S-73-1, Report 6, 1977 .

25 26 ~,. D. B. Slemmons, "Determination of Design Earthquake Magnitudes for 27 Microzonation," Proceedings of the Third International Microzonation 28 Conference, Volume 1, pp. 119-130, 1982. 29 30 -51. M. G. Bonilla, H.A. Villabobos, and R. E. Wallace , "Exploratory Trench 31 *Available for inspection or copying for a fee at the NRC Public Document Room, 32 2120 L Street NW., Washington, DC . 24

NUMARC Comments March 19, 1993 Line In/Line Out 1 Across the Pleasant Valley Fault, Nevada," Professional Paper 1274-B, 2 U.S. Geological Survey, pp . 81-814, 1984 . 3 4 ii. S. G. Wesnousky, "Relationship Between Total Affect, Degree of Fault 5 Trace Complexity, and Earthquake Size on Major Strike-Slip Faults in 6 California" {Abs), Seismological Research Letters, Volume 59, Number 1, 7 1988. 8 .09 11

   +1'0.. D.

-:-:====**:

L. Wells et al., "New Earthquake .Magnitude and Fault Rupture Para-meters: Part II. Maximum and Average Relationships" {Abs), Seismological Research Letters, Volume 60, Number 1, 1989. 12 13 S:~! 1{ . M. Wyss, "Estimating Maximum Expectable Magnitude of Earthqua*kes from 14 Fault Dimensions," Geology, Volume 7 (7), pp. 336-340, 1979. 15 16 9:1:2. D. L. Wells and K. J. Coppersmith, "Analysis of Empirical Relationships 17 Among Magnitude, Rupture Length, Rupture Area, and Surface Displacement" 18 {Abs), Seismological Research Letters, Volume 63, Number 1, 1992. 19 -~ 22 23

   ~J:'-.*        Letter from G. Sorensen, Washington Public Power Supply System, to USNRC. 

Subject:

Nuclear Project No. 3, Resolution of Key Licensing Issues, Response to Question on Seismic Hazard, February 29, 1988.* 25

  /Jl/6£3

~6-rr;;;_7 I/~£ p1w1ee:s

 !+~

NUMARCCommenta Site-Specific March 19, 1993 Regional and Site Une ln/Une Out Geological, Seismological and Geophysical 1 Investigation 2 EUS wus 3 4 Perfonn Integrated Develop Seismic 5 Evaluation of EPRI or Sources and Seismiclty LLNL Seismic Sources Parameters 6 7 8 No Revise/Update 9 eo11 Seismic Sources 12 13 14 15 Perfonn Probabilistic Seismic 16 Hazard Analysis 17 18 19 Determine SSE Ground Motion 20 tl1 22 23 Determine Site-Response Spec::lrUm 24 25 or 26 J 27 Scale RG-1.60 Spectrum Develop Site-Specific Spectrum 28 29 30 f;gure 1. Flow Chart for Determ;nat;on of the SSE ;A the EasterA UA;te~ 31 States . 26

                                                                                                                              ~ Comments March 19, 1993 Une ln/Une Out 1                                         Sita and Site Region                        EPRI Geological, Geophyllcal                  Earth Science 2                                           and Seilmologic:II                      Database IIMlligallon 3

4 5 6 EPRlorLLNL Seismic Hazard 7 Analysis and Information Base 8 9 eo 11 C i EPRI orLLNL Seismic Sources, Selsmlclly Parameters, and Maximum Magnitude 1w 12 0 (I) 13 1 EPRlorLLNL Seismic Hazard 14 ~ Analysis and Cl Information Base C 15 1 Develop Seismic Sources, Seismicily Parameters 16 j and Maximum Magnitude Based on New Data 17 18 19 20 j Seismic Hazard Analysis and Information Base 91 22 YN 23 24 25 26 Site-Specific Update of Seismic Sources 27 28 Figure 2. Schematic Aepresentatien ef the Qeterminatien ef the Centrelling 29 EarthEtYakes fer the Qeterministic Analysis Path. IIJfflltliiI\P:l::Iill .,.,.*.*.***********...*.*********************************************-************************** 30 §.l:J~SP.d:t:li:s}]:J.Htlt.g)fa.t.)otti1Rib¢e.:$Jj::;f;ftlt{lvilfigl.iJhlii\J.fi~t:fEMl'f:o.t@l):.~N~ 31 ~,11:1mll.:l$.iU.tl.i.it:::llt=mm:tl1.it.i.fflW[Ui::JT.l 27

NUMARC Comments March 19, 1993 Line In/Line Out 1 APPENDIX A 2 3 DEFINITIONS 4 5 A capable teetenie seuree :li.lH:I is a teetenie stl"uetul"e li.iJl:t that eaJt 6 genel"ate ~eth eal"th~uakes and teetenie sul"faee defePmation such as faultiRg or 7 folding at 81" neal" the-~~ suFfaee iR the ,l"eseRt seismoteetoRie regime. 8 It is eharaetel"ized l'>y at least oRe ef the follewiRg ehaPaeteFisties? iii:1,:1,:1:, :1 9 111::iett:mr1::rei:111l:r111::1::,11it1:1s11n1itit1:11:1si::ri: 10 e112 a. PreseRee ef surfaee er Rear surfaee defermatieR ef laRdferms er 1:i::il)iii.llilfiP!I geo l og i C depo s i ts i.lIIiit:::1;,1:1:1111:tsilinl:t:11111,i. 0 f a 13 recurring nature with i n the last approximately 500,000 years or at least 14 once in the last approximately 50,000 years. 15 16 b. A reasonable association with eRe er mel"e large eaPth~ualtes er sustained 17 mlllrili:/::::i1XIJ::1m@::::::1:1r1iiYi:~g§I::1n::::::11§li.li@g earthquake activity, usually 18 accompanied by significant surface deformation . 19 20 c. A structural association with a ~PP!! capable teetenie seul"ee li~Hll 21 having characteristics of m~i@!!li]fa;),! of this paragraph, such that mevement - 23 24 25 llli!PliliffllfJ!I, on one could be reasonably expected to be accompanied by movement on the other. In some cases, the geologic evidence of past aeti'lity ~!jl)i im!S!'-ffl!!i[I at 26 or near the ground surface along a particular capable teetenie seuree f:iimt  :*:*::!:-::*::-: 27 may be obscured at a particular site. This might occur, for example, at a 28 site having a deep overburden . For these cases, evidence may exist elsewhere 29 along the structure f:t.i.lil' ffeHI Pi which an evaluation of its ehal"aeteri sties 30 !liiiliifiliillir.i in the vicinity of the site can reasonably be basee ;,::n iirriP

• 31 Such evidence should be used in determining whether the stFueture ~iiUll! is a 32 capable teetenie seurce within th i s definition.

33 34 Notwithstanding the forego i ng paragraphs, structural association of a A- 1

NUMARC Comments March 19, 1993 Line In/Line Out 1 struetl:H'e liimt with geologic structural features that are geologically old 2 ( at 1east pre-Quaternary) such as RlaAy ef those 11:iiriliii found in the easterA 3 regieA ef the YAited States HI will, in the absence of confli cting evidence, 4 demonstrate that the struetul"e li.1.J:1:] is not ,1 capable teeteAie seuree within 5 this definition . 6 7 GeAtl"elliAg eal"theuakes (GEs) are the earthquakes that preduee the 8 largest estiR1ated greuAd R1etieAs at the site. There R1ay he several GEs fer a 9 s-1-te. ea 11 A deterRli Ai st i e seuree eal"theuake (QSE) lb1:: : ~~~:~~~:: :m iini:IJ.is~:: :=~~!~li~il.l 12 is the largest earthquake that can reasonably he eMpeeted te occur in a given 13 se i sm i c source 4ft iin:,:1:111nl:II11:11:r11111:1,:j:e,1::i::i:11:j;J:is.1i:silIIilt:::11111111:1 14 i!limilliiJi!!ll~§i;::::11t:::::~:n1:ts:1rrt:n1::::::1gi:llni:i regi me[~: , aAd 1t is used i A a 1 15 detel"RliAistie aAalysis. It is geAerally hased BA the R1aMiR1UR1 histerieal 16 earthquake asseeiated with that seisR1ie seuree, uAless geelegieal evideAee 17 waFFaAts a laFger earthquake BF the rate ef eeeurreAee ef earthquakes 18 iAdieates the likeliheed ef larger thaA the laFgest histerieal eveAt. 19 20 The iAteAsity ef aA eaFthquake is a R1easure ef its effeets BA huR1aAs, huR1aA hui 1t struetures, aAd the e'.l arth' s surfaee at a parti eul aF 1eeati BA

• 22 lAteAsity is deserihed hy a AURleFieal value BA the Hedified Herealli seale.

23 24 An earthquake magnitude i s a measure of the strength of an earthquake as 25 determined by seismographic observations. 26 27 Nontectonic deformation is di stortion of surface er Aear surfaee soils 28 or rocks that is not di rectly attrihutahle te 119.il:i~]!iil tectonic activity. 29 Such deformation includes features associated with subsidence, karst terrane, 30 glaciation or deglac i at i on, and growth fault i ng . 31 32 The sl afe sl hutdown ej arthguake al round mfflotion (SSE} is the vibratory A-2

NUMARCComments March 19, 1993 Une ln/Une Out 1 ground motion for which certain structures, systems, and components are 2 designed to remain functional. 3 4 A seismic source is a general term referring to hath seisMageRic saYPces 5 aRd capahl e tectaR i c saypce s 1!99'.!IPl:i:s::::::111:~:1#1::::::!:11~1::1r:1:::::::1::1~1!!iin:11::e:1::::::1,1;§1J1 6 @iillfiii.li:I:isl:1::1:i:ii:I:liv:i:,lii!:liil!J,:n,ll:iliniilinl:1:1:1:sii. 7 8 A sejsMaeeRic saurce is a partiaR af the earth that has YRifaPM earth -~ 11 12 qYake pateRtial (saMe eMpected MaMimYm ea,thqYake aRd fFeqYeRcy af recYPreRce) distiRct fram the SYFraYRdiRg aFea. A seismageRic saYFce will Rat caYse SYF face displacemeRt. SeismageRic saYrces caver a wide raRge af passihilities fram a ~*ell defiRed tectaRic stFYCtYFe ta simply a large regiaR af diffyse 13 seismicity (seismatectaRic pFaviRce) thaught ta he characterized hy the same 14 earthqYake PeCYPFeRce madel. A seismageRic saYrce is alsa characterized hy 15 its iRvalvemeRt iR the CYFreRt tectaRic Fegime as reflected iR the QyaterRaF;Y 16 (appFaMimately the last 2 milliaR years). 17 18 A stable continental region (SCR) is composed of continental crust, 19 including continental shelves, slopes, and attenuated continental crust, and excludes active plate boundaries and zones of currently active tectonics '22 23 24 directly influenced by plate margin processes . It exhibits no significant deformation assaci ated wi thj!Ji the ~ Mesozoic-to-Cenozoic (1 ast 240 mil 1ion years) eFegeRic helts J.!~~Jili'=i~- It excludes ~ zones of Neogene (last 25 million years) rifting, volcanism, or suturing. 25 26 A tectonic structure is a large-scale dislocation or distortion YSYally 27 within the e§arth's crust. Its extent is on the order of miles. A-3

NUMARC C,omments March 19, 1993 Une ln/Une Out 1 APPENDIX B 2 3 ACCEPTANCE CRITERIA FOR THE ANNUAL PROBABILITY 4 OF EXCEEDANCE LEVEL FOR SAFE SHUTDOWN EARTHQUAKE 5 GROUND MOTIONS 6 7 8 B.l INTRODUCTION 9 10 This append i X 8Ytli Aes 111,:r:Uli: a procedure to cal CYl ate llli1!1i,1 -the 11 11::::::11s111:111,:111;:1:1:m~:inII:tii[:llni annual probability of exceeding the safe 14 shutdown earthquake ground motion {SSE). 1:!:1:,11::1tl]IIJI!il:1n11tJ;ni::llmli!li lil:il!ilU:lili:IIl.li]:lilillilil:]lriii§:~:m:tlilil:iill:lliIIIIIJ:IIlbill:iilillil.till:!l:1i::!Iiil:imls i1111r1:111:11:1:,:i:1:r1:11IIBIIil111:r11:r11:1:iiffij::n1::r1:111::11:;::1m1:it]ll1J:111:111:1:111:11:11::1111111, 15 11::111:::::::1:1:11:~:::::::1::11::::::rt§ill:liilIIfi.iIIiiiiilli:iI::1::::::11:11:1:Is!IIIBiil!Ia:11:~IIlliI!llllii!l21 16 :1telt:11:m111=r::1=::t'lll!f}l!III::liitfil:tlt:11:1:=::::l!IIIl!:l!IDl£iilIJel!!!ilIIl!ll!!iiS!m!,rl.! 17 mi1ir1IIIl!IIlitilhill:iiiiliallliiiil:I::11::::::siiriillii:!I!illiil!ll!niI!l!il:lii!iil:li!rt::::ilii:IIIIR[li!i!I!::,1 18 l[l§Elisl:~:11::::::111:11:1~::::::111::~:::11~:111:l::111:il@ This pfeceEIYre caA he YSeEI c1) te 19 ce~pare the calcYlateEI aAAYal prehahility ef eMceeEliAg the SS~ at prepesee 20 plaAts te the calcYlateEI aAAYal prehahilities fer the cyrreAtly eperatiAg 21 plaAts as reqyiree hy AppeAEliM 8 te 10 GFR Part 100 aAEI (2) te estahlish 22 ceAtrelliAg earthqYakes iR the prehahilistic hazare aAalysis as Eliscyssee iA AppeREliM G te this regYlatery gYiEle. YAiferm hazare spectra (spectra that ha¥e a YAifer~ prehahility ef eMceeeaAce e¥er the freqYeAcy raAge ef iAterest) 25 sheYlEI he calcYlateEI te estimate the aARYal prehahility ef eMceeEliAg the SSE 26 ElesigA respeAse spectrYm. 27 28 1:l:1: : : : : : : : : :~~!:~~~~!:~: : :~~!~~'~:~:m~:,~~:: : :! :~r:~:;:~!: : ,:~~~ 29 30 111:1:11:1::::::::1:1:tliiilUli!II!::liinHiiI!::iiiliit::1:1::11:i1:1:i:1::1::,m:i1:11111:rlil!liiliill1 31 :1:11111:::::::11:111::::::11::::::11:I::1111e11nsil!IirU:llii!iiRl]lliilIIl!lillilnilli!Il!lll1:iiI::::1m11.1!:l]::1:11:11:1 1 32 !iiiffliilill:l:lllllalit:l{iliiilii!li:tiirl1!!!llin::~i::111:1:,11::::::1:,t::1,1i1t:11: f:ii1,11:::i::1iili:Mi:i::,:11:i: 1 33 :i1i::111::::::1:11111i:1::;:111:1iiill::!i:isfiI!IIHlitIIIIIIIillilltii:Iii!trinl:J:lilli1111.t:u~1:l:iJ.:ilil1:l:Il1i1. 1 1 34 illliiliii:J!liiil1lliiIIiiiiil1:m~::11:111:Iiiiiil1:,1:::::::111:::111:11:1iliB!liill1:[!l!lltU:1::i:r:irn11 1 35 liiiiinii'ijliliili:J::,:11::::::;:1:11,11:r1:i111:::1111::::::1n1IIinnilillIIiilliilJ:!:illxIIi:IIIiisiilllll B-1

NUMARCComments March 19, 1993 Une ln/Une Out 1 1:11IIlliiinlil:l!Pii:i:1iffl:fim:Hlili:Jiiiri:11111,::ns.]:li:tlii:J1!illii[\g1IIIIIiilIIl:11:11i: 2 3 There arc twe state ef the art approaehcs (eutlincd in R&fs. 18 and 28) 4 eurrcntly a¥ailablc te ealeulatc the prebabilistie scismie hazard fer sites 5 cast ef the Reeky meuntains (Eastern Ynitcd States). These appreaehcs, hew 6 c¥cr, produec different hazard estimates for a gi¥cn site. Therefore, the 7 NRG staff is rceommcnding the following interim proecdurc until the 8 diffcrcnecs bct,iccn the two hazard mctheds arc rcsol¥cd. This proecdurc 9 relics on rclati¥c measures to ensure that the annual probability of cKeccding eo11 the SSE is eomparable to that of operating plants. The proecdurc is based on studies eondueted for the Eastern Seismieity Issue and the IPEEE program (Ref. 12 38). Either the LLNL or EPRI methodology ean be used to earry out the 13 following ealeulatiens, ,iith the appropriate set of limits assoeiatcd with 14 caeh method. If any analysis other than the LLNL or EPRI methods is used in 15 the Eastern Ynited States, annual probabilities of cMeccding the SSE would 16 need to be dc¥clopcd fer all operating plant sites in addition to the site 17 under eonsideration in order to make the appropriate eomparison. 18 19 fillll!tEilm=lr:=J:J::lit~l!ilt:!ililfit:11:t:§:!ni::!tl:l11rn}lltlli:1n::re:m1111:m:t!J:@tI:!Rl}?lli!IIJi:PQ 20 1:111:111IIilitttei:,,1:,::1:,llri!t!i:!ns.I]i::1:i:t:ijil!til1lllit!e§n!!i:iinli:)Itllpn,:::1,:1;:11::~::::::::11111I:1:1n - !n1:1111:11::11.11lrt:tn@IPJ::11@1:1:wl@:1:111triis1II!nili§!:mi1i:1:t:!:I:IIIllIHfif 22 23 B.il PROCEDURE 1':tif=DEfflERtr::-****.)(:*,;::m::ss; 24 25 The folle,iing proeedurc is one approaeh aeecptablc to the NRG staff to 26 assure that the annual probability of eKeecding the SSE eomparcs fa¥orably 27 with that for the nuelcar power plants operating as of the date of the final 28 *1crs i on of ~ppcnd i

• 8 to Part 100. J.iti@Jlj1J:lqJ,1jJj,JUti.!lJ::::119!:!~9il!i!ljijljj\1,jljl1li!1IUJljj,g 29 lg¥:1:rm1:n1:1111:::::::11:§::111:r:1r:1:1::1:~rn::~:H1~r:i:a111:1::r11.1Pr1R~:>.m::1~:=~mmp;~~um:::t=;111:m:1b.1 3o !M!Etn!:1<<t:9P!J!!i:i:n::::::nl~9!:1:1!'.t4¥l¥!!J\\P1:tnliti:1lttliltt:::=§xl::c,:E::;:p:~::1[:~f;f111:;::1@t:~1!s:tef 31 1:11I:::1~::11Jil!illill::lf 32 B-2

NUMARCComments March 19, 1993 Une ln/Une Out 1 WHe.:rtt=tfie.t:ap:p.J:=tc:ant.trus:es:t:a=:rPS.HArme.t.flqpt:p.gy::tt:ha:t:Jthi'iitbeen=:bicceptnr:1,-:rtt11 2 Qffili:§."l,ofi,@Uti~sMt,l:I.UtZBJ::l'f,thlftNRO.:tstif;ftw]=J:it{geff.e.tall:yf:p.o.nt,tf~iti.tlflj:t\f,)}:t.eMt!.!I 3 o.lI;!late.t11:iilil::mi:it\:lthe.ltilt.tio.d1Ui9¥Wfiii:::\::ti1tttl1imnt11tt:J:at.e.J.1:tr1run::t'-1:mi1l::rntb.ew::iitpul. 4 11,ra*ter:,:t=1rg=@at.t.e.pt.:ablewm::=1:tv1nm::\1.e.ol:gg:t:e.11:Mt1e.oeb.v:$;i:p1mw:=:=1n1.:r'1,u:s.}:no:1:p.g:J:p,1l: 5 tt1iiddevelope.d:(l:'.or=tt.fle+snc.= UM:crts:tt.e.t:under:f=:~v1ew.v:/+ini:11pl::i:c.= antemaktusei11 6 ISHA=::::iwt.H9\10:lo.Q1not.=:::::1.11vi:0:,'-'l:Yt'\4'¢@1itil::::::tix::rli1:1lt:1if@trtl~n1tt.fi11t:1=i.\ltn1l:l@n1:,:: 7 lmm$J.1ff:lt@1\li:J:=tt1t1:,w:t~b.@\HIPPtllnt1=:111n.1.1,r::1f::}!th,::ni.i@~hB.P:JJ1.gy1:1nl.tAf!.J 8 DP!::l=!i!J!f:t?eni!i1t:s:lili:!Plsit:ll&f:li\!l:t:1@ 9

• B* ~ ~ , 1 l.lill=1i.: =: :!!.nl:i:fiii~:i~:: : ::~:!:#. Ea steFA 1:1. S. Si te s 11 12 FbJM='l=i:J:$?::l:n::::=:t.tig=:::::SCR~::JJ¢¢1Pt.ed:::=::s:e.:lll!ift.mr~P(0f:¢e.::rnitilltP:tte.t.=a1:ua.t1t:mp::::::fi 13 i.1$.e.d:tpj(lipp)::J:¢.ib.~j:\\:]w.fit.fii:H:fijj\l#,=liftn,:rna,1,rm:1:b:tpg@ifie.'{£$1\ti#.f::j})~=i;tiil(ilfWlb.it:$1.Ri:

14 f.1)1l\i.s:s}t1:11:11:1.¢.r:Ui¥:4.:::::::11:m::u.;1f.o.d1\:Kfi=r:::::=t.11;v:m,r,t\i.:b1Mtltiltil.=\\j\\:jp¢\1tt1.l:$.l.l,ri.ltNdtifi. 15 s,l::f.e.m:*ee.e.1:ri<<raa:t:a.:~:t=t:::ifijj:se.=1s.Jtes.1::re.qubt1:'ts.i=tst::1µe.e.i.:J.fillc.,is.e:1::s.mi::e.qr.so:urt.1+::and 16 :11*I11::::::11n1m1:11r::::m::n11:1111111:1:1n:1m11::ri:11::::::1i1:i1:::::::1:9nJI1111rmi1:i1s::i:::11,111111:r 17 18 Step 1. li:J.:n§l]~n:1:1~pgjjjj]::es1.1:r,1Jt:111,1::1s1i:1:9.1)':ilH~t.l:lthiil:l~H:IHi::::m11irl 19 l!f:9:ll!!i\iillL~t:;:::~:fi!lI:i:J::1,1::~: Gal cYl ate the Yni fePRI MazaPEi Respense 20 SpectPa (YMRS) with vaPieYs retuPA pePieEis. FiguPe B.l shews a saR1ple set ef R1eEiiaA YMRS feF vaPieus retuPn peFieds. The YMRS 22 sheulEi he Eieveleped at the saR1e lecatien as the lecatieA ef the 23 SSE ( i . e. e i theF at the fpee gPeund supface eF at *a h:)'pethet i cal 24 F8C k 8utcFep) . l@~::1m~:1::::::1:11it1::::;p,1.ittv@i1::]ili9]::~::::::,.::::::1,1::si::1::1;iiUi::~11 25 ifiiil!nil:!IIiiiil:iiil:Uin::::::::(:l~').iliililil.@:~i:::1:1:::::::1J::1J:::1111:::::::11iu1@ 26 27 Step 2. 1:1:i:,,1::111::::::111im11:l1ni:1:1wm1:i:11:::;:111111::::::::1m111,:1::;:11:::::,,tm:1111111e, 1 28 ~t,Eli1e.ptr1J.:]:=ic..i1liir.i.li bn:,::::::1~:1:s.:~:01:::'inli]l1Q\:~:11::11IIinl!.ll.@O.lliil1Il1llllii. 1 29 Galculate eeR1pesite annual pPe~ahilities ef eMeeeEiing the SSE and 30 eeR1paPe these pFehahilities with eperating plants using R1edian 31 hazard estifflates. (Altheugh the R1edian estiR1ates ape useEi feP 32 caPP:)'iAg eut the pPeceEiYPe eutlined in this appenEiiM, the hazaPEi B-3

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 analysis sheyld he perferMed with eensideratien ef Yneertainties 2 te develep eeMplete insights.) The preeedYre is illYstrated in 3 Figyre 8.2. 4 5 [stiMate the annYal prehahilities ef eKeeeding the SS[ 6 speetrYM at twe diserete fre~Yeneies (§ and 18 Hz) Ysing the 7 YHR-S. 8 9 2. Gal eYl ate tfflhe composite annua 1 probabi 1i ty  :!:1:tl'=!!!>>.J:n.l,t, eo 11 by using the fellewing formula: 12 Composite Annual Probability= 1/2{al) + 1/2{a2) 13 14 where al and a2 represent median annual probabilities of 15 exceeding SSE spectral ordinates at 5 and 10 Hz, 16 respectively. 17 18 ini:II2im;9J:,:1:1:tinni:11t::::11ili§l:JI1::11t::1:1t:iiliil:Iilili~el:iilit:niifil!lil 19 lil:Iilit::1:luig 20 22 23 [KaMple: FreM FigYre 8.2, fer a Median YHRS derived Ysing 24 the LLNL Methedelegy, at paints al and a2 eerrespending te § 25 and 18 Hz: 26 27 GeMpesite AnnYal Prehahility . 1/2(4[ §) 1 1/2(8[ §) 28 - &[ §. 29 30 3. FigYre 8.3 shews t~e distrihYtien ef Median annYal 31 prehahilities ef eKeeeding SS[s fer eperating [astern Y.S. 32 plants ysing LLNL hazard estiMates. This figYre alse 8-4

NUMARC Comments March 19, 1993 Une ln/Une Out 1 indicates a limit; appFoKimately §Q~ ef the CYFFently 2 epeFating plants hat.*e an annYal pFehahility of eKceeding the 3 SSE gpoynd motion helow this limit. (Limits ~op hoth the 4 CYFFent EPRI and LLNL seismic ha~aFd stydies aFe listed in 5 Tahle 8.1.) The SSE is adeqYate when the annYal pPohahility 6 of eKceeding the SSE compaFes fa¥oFahly to the limits shown 7 in these figYFes. 8 9 11:ii{Ie:g:11::111:1111111:P:l!I:lliill:§:§t::1218'.9:!i!!liitlir[;:t:nt:I:i!@!tl.!iiP!f:iJ:1::::::111r:11i11::11:i@nl: ea 11 iJ]:lf!lil::J: 1::~::1:JliI!Iililliini.Il:J:lnifiiil:l:1el:l:l!lliilnlI1iiri@illlil:ltl'=lnill:I!ilrnl1'.ir 11 l&IIl:iliii!fi!iI:/:Ii:iillllllffl!l§l!t:11::::::1n1!1:Illlirill!IIlillliiJii!lii¥::mi:rn1:1:1i:1:1 12 :li:l::llfi:Jntto::J:IlblitJ19.iffiillitlJ: 1 13 14 step i I- l:li!flilfl1:i1iniflnii:ienlili:l:liiiiiliili::l:l]l:il:l:l:lll@:rmli:iiiflli!I!:11ii:liEiliri 1s 11111i:111::::::::~:i::::::11ii:11:i1::i1:l:[:i::rn1,i:::111~t111:1,: 1111:~:1,:1::::::111:::::::111m1::111:i:i11:1 1 11 16 ii1:1:11::::::11t::1:ii!::1::::::1::::::;,:1:liiiiis:1:tj![p!:i:Jiiiiiinlii:i,:1:1111it!liilii,U11iI::::1 1 17 ilinliilI!IilieillI!lllfllilllil:IinlIJ::11r:,1:::::::1.n1i:111:11::ili:Iinlt::1i:1:::::11E:iiislriJ.l. 1 18 issilllitl:1::en:::::::1::111:l::1::::::111111:11111:ni.Il!l:illllillliliiiiiif!miliii/ll:1:11 1 19 :1:1111::::::1111::::i:1:):i:l 20 B-5

NUMARCComments March 19, 1993 Une ln/Une Out 1 Tahl e 8.1 2 3 4 Method Annual Probability of Exceedance Limits for Median Hazard Estimates 5 Y:Nh -I-E-4, 6 EPJH. 3 f § 7 8 9 FeF the hypethetical eMample, the cal~Ylated aAAYal eo 11 pFebability ef eMceedaAce ef 6f § is less thaA the limit ef If 4, aAd thys the pFebability ef eMceediAg the SSE cempares 12 fave,ahly with that ef epeFatiAg plaAts. 13 14 FigYFe 8.4 preseAts the same iAfermatieA fFem the use ef the 15 fPRI YHRS estimates. This limit sheYld he used wheA the 16 fPRI methed is used te calcYlate the aAAYal prebahility ef 17 eMceediAg the SSf. 18 19 B.-23.2 Western U.S. Sites 20 91 For the Western U.S. sites, a probabilistic data base, such as that 22 compiled in the LLNL and EPRI studies, is not available. Te date Ae pFecedure 23 eMists te cempare the aAAYal pFebability ef eMceediAg the SSf teether sites 24 iA the WesterA YAited States. In addition, the probabilistic hazard at a site 25 in the Western United States may be governed by /jg)j.iJl.1[i;!i)i/JgijJ1ii: clearly 26 identifiable seismic seYFCes 11::::::11,1:J::::litlnitilii:liiYtlis,, such as faults (or folds) 27 ebseFved at the syrface that have better defined seismicity !iPI!! 28 characteristics. Therefore, for Western U.S. sites, a site-specific analysis 29 should be de*,eleped YSiAg suitahle methedelegies i!IIU~J!ij!!i~i! to estimate the 30 annual probability of exceeding the SSE and to identify significant 31 contributors to the hazard (Ref. 4B). 32 B-6

NUMARCComments March 19, 1993 Une ln/Une Out 1  :~ltf:tm:U11fi:!)p],~Mtr~h~fNIC9q~4r~:1:~~v::fl'"~'J'Sll!!Wifi@ti§~lt=il?i~r::~:~~1r::i=:rr@t.h~trWt,$.,llm 2 Y:tl-Ar::~..1::l~f.@'~:b.lfJtll~?=l:H,::::::il~W\'f::::,111vt.JM=?\:P:P.l:llPt:lm~f@l-lPl!fJf!:n:::111,ftiJ.11,ng1ti.l1D 3 ~~t:1:tto.ngr:m:t:to.ntMiJrttte.:tEaitertura=tl1J.tennn=1:ts.Mwrtba.:n11:>>1en=*11ib.uat1'rtunves 4 !tiiiltbe=tWil'lllttt}U.Mi=Wflittdt::t:u.:net=lt~llJtrthllit:ttio.ll'liltlN.ftil=,El.$.IJt:f:l$!HJ.'tt111.UJ.ili= 5 lhit:lifi.ti.btit'!Pr.6.&iJ.U:l4iJ:ttl\tie.t.erffi.i=ngt\t.li\J$$11!1tl=i\i'tgJ:,jli!n=lfifill!Will.to.t:U.il.M 6 **¥tlit:fd:1ffit@mttli11Df=Pll!flt:¢.ffl!ljjf\il'd¥t.l!l.¥t.l!U1til=iit'ttf.Uml'i6\ttlli.l'rt.!tlhltll.tt\llfh:

. 7 1:t11:111.it]iilJ::1::r1111:;:,11::::::11,::::::mt:i:1:1i::1:1::::::111::::::11:1:j!!iiliirfili!ll1i,li:I1.:i111111,i:1 8 11111:11:mi:11:::::::11r::f:1:11:i11111:1,::::::11:1:;:

-~11 12 REFERENCES

18. Electric Power Research Institute, "Probabilistic Seismic Hazard Evalua-13 tions at Nuclear Power Plant Sites in the Central and Eastern United 14 States: Resolution of the Charleston Earthquake Issue," NP-6395-D, 15 1989.

16 17 2B. D. L. Bernreuter et al., "Seis~ic Hazard Characterization of 69 Nuclear 18 Plant Sites East of the Rocky Mountains," NUREG/CR-5250, January 1989. 19

38. J.T. Chen et al ., "Procedural and Submittal Guidance for Individual Plant Examination of External Events (IPEEE) for Severe Accident 22 Vulnerabilities," USNRC, NUREG-1407, June 1991.

23 24 48. USNRC, "Safety Evaluation Report Related to the Operation of Diablo 25 Canyon Nuclear Power Plant, Units 1 and 2," NUREG-0675, Supplement 26 No. 34, June 1991. 27 28 8-7

S5J;J 22/fr'

              \,

l111_dp

   ~.9~

NUMARC Comments March 19, 1993 Une ln/Une Out 1 2 3 4 5 6 7 Composite (5 and 1o hz avg.) of Exceeding 8 Design Basis Using EPRI Median Hazard Results 0.001 . - - - - - - - - - - - - - - - - - - - - - - - - - - - - , 11

    >-0.0001 12 ~

Median =* = 2.0SE-5 " " 13 .0 14 ro 15

   -g.... 1E-05 16 a..

Q)

   ~

17

    ~ 1E-06 18  a.

19 E -~ 22 23 0 (.) 1E-07 1E-08 ~----'----'-----'--------------'---'---...________._ __.__ __.___...________._ _.....___, 24 0 10 20 30 40 50 60 70 25 Sites 26 27 28 29 Figure B. 1 Medi aA UFI i farm Hazard Aesp8FISe Spectra ,:,;:1::u:i.£.rit:J.ln['b.lfil:::::1n1 30 Pro.ciaure=r:to.JD.etermhnrr:ttiet'Re:f.erencer=:Pto&.as:t1dttJtf.ar;f~1:~~skltiN~~~ 31 ltR:::Us4ngithe'tEPRl:tMedian\Haz:~~~,,:,~~~~~ ~~ B-8

NUMAFCC,omments March 19, 1993 Une ln/Une Out 1 2 3 4 5 6 7 8

• 1.00E + 00 1.00E-01 11 --o-- 10 Hz 12 t .00E-02 - - - - - - - - - - l *---o---* 5 Hz 13 - - o 10 Hz 14 1.00E-03 15 1.00E-04 16 17 1.00E-05 18 1.00E-06 19 20 I e122 1.00E-07 1.00E-08

                   ---***-* ---*-************-*-- -*------*-*-*-*-*****--  -   *r- *-*--

1

                                                                                         - - - - - --     - --=~-

0 100 200 s. (RP)._10 400 500 600 23 Spectral Acceleration (cm/sec/sec) 24 25 26 27 28 Figure B. 2 G8Fflf)Yted ,.AAYal JU\oeelu.r<<mtJi?D.if!.t.ffl\i:Jje.:tt.fig:tcapo.:~mte Probabi 1i ty 29 of Exceed~a.ff.ie GesigA Basis 30 G8Fflf)8Site AAAYal PFehahility l/2(al) I l/2(a2) B-9

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 2 3 4 5 6 7 8 9 ~ 11 12 13 14 15 16 17 18 19 20 e1 22 23 24 25 26 27 28 AAAYal PPohahility of ExceediAg SSE 29 30 Figure 8.3 AAAual PPohahility of ExceediAg SSE YsiAg MediaA LLNL HazaPd 31 Estimates B-10

NUMARCComments March 19, 1993 Une ln/Une Out 1 2 3 4 5 6 7 8 9 eo 11 12 13 14 15 16 17 18 19 20 - 1 22 23 24 25 26 27 28 29 30 FigYl"e 8.4 AAAYal Pl"ebability ef EMeeeeiAg SSE YsiAg MeeiaA 31 EPRI Mazal"e Esti~ates 8-11

NUMAFCCiomments March 19, 1993 Une ln/Une Out 1 APPENDIX C 2 3 DETERMINATION OF AfSElSM:MNJJAZAltfJINJfORHATfON!JBAS.E :'FOR 4 IAEE ': IBHDnwtt:'IIBiHDfffiKE<IBDUNl<M.QJ'.HlN .. .. ***...* ** 5 CONTROLLING .EARTMOUAKES FROM TME: PJmBABillSTIG . ANALYSIS 6 7 8 C.1 INTRODUCTION 9 10 This appeAdix eutliAes a precedure te determiAe ceAtrelliAg earthquakes ei frem the prebabilistic hazards aAalysis fer a site. The greuAd metieAs frem 12 these ceAtrelliAg earthquakes sheuld be determiAed fellewiAg the precedures 13 eutliAed iA SectieA 2.§.2 (SubsectieA 2.§.2.6) ef the StaAdard Review PlaA. 14 GeAtrelliAg earthquakes sheuld be determiAed fer the mediaA seismic hazard 15 limit used ta satisfy the requiremeAt ef AppeAdix B ta IQ GFR Part IQQ aAd 16 discussed iA SectieA G.2 belew aAd AppeAdix B ef this regulatery guide ta 17 demeAstrate that the aAAual prebability ef exceediAg the safe shutdewA earth 18 quake greuAd metieA (SSE) cempares faverably with that ef the curreAtly 19 eperatiAg Auclear pewer plaAts. 20 21 J=td:s.t=yperil.ilt:Ulleib.hi:b.i:~rliJ:pro.c.e.4ihf:e.:rt.o.t:dit.l.rii:rte:tt.hi:t:s.:ife:t:s.hut.do.wti a 2 23 iilltiiiiliiii:r1iill[!lljJ§i:1::1:1111:r:111:1:J:jjJJillinl::::::11::::::1111!J:11:::::::1::::::11,jJJiiiliJ::1:1:,,11

i:nl:9!rm!ilj::1n:::::::§111mr=:rmn11::l:n£9rm,i:~Jm:11111:1:~Mi!!.1r:~:!¥:!1:1:n1t::s1niw~::1µ~M::~mr1:t 24 1=:n1:1::v,:11t.:il:t':jgjjjffijUh):)$.9lttl.l)fi.e.::umit'ffiiilUl.ude::tind:::=:fJli!ti6.tlflb)t:tiirlHqta;ie.:s.:::=:::t.ha.t:

2s a1~11te.=: :::1He::::::s11:1,n:c.:ta1s ,:111:11v11::w::::::::::::11.r:1:1:1111,1::b.1R:t1ae1111::::::1s:1:::: 1aw1e:s10.naJJ1gt:t.o. 1 1 1 111 26 :1H1::::::a,r.*11.ni<<:r:eJr<<§itMrn:1::t&::::::111:J:11.tt:::::1:nr::11un111'x1:1H:::11::::::1:n:J.:1:m111u11M:::::::mr,,::11"e.m:r1:$. 21 1.1timt1illiit111mii.liPi!l!!~::1::1:1:,1:stI11:l1m1:s:li!l!!IP)l:i1n1l:x,:1:1ti::t11Ml:!t:Ii:::l:1::!I1:~:i!II:i:! 28 1:11:1:1.:tlllrlli.U.iilI:i.pie.lB.'Iih)iii::rn::::11:111:rt@i:ii.W¢i:fJjp)lr.ii.POJJ)$.)l::::::1p.19.);:i~ijijf{ihiJAH@i::, 1 29 llllfmUjlJ.j:jjj::f§IJitb.if'U.ibl)ffli91'i::111~u::1111:1m111ntillll:1nlJ::1:1e.1:1:::t$.!i:J:lfi¢.{Uil:irnt:i:itti.MJ 30 111111:1:1::::::1:tl!iI!lli::1:::::'IB.!l::ili:t.f:t:::1;:j:91::::::11:tl!:Ul!l'l[:1111111me::11::::::1111]f[ll)!!)!l!P)!I 31 :1:11i1t:Mi!l]lli.r!ifflif!lli]lllll: 32 33 C-1

NUMAFCC,omments March 19, 1993 Line ln/Une Out 1 C.2 PROCEDURE 2 3 The fellewiAg pPeeeduPe is eAe appPeaeh aeeeptahle te the NRG staff te 4 detePMiAe eeAtPelliAg eaPthquakes fpem a prehahilistie hazards aAalysis. 5 6 G.2.1 EastePA Y.S. Sites 7 8 As diseussed iA AppeAdix 8 ef this regulatepY guide, thePe aPe twe appreaehes (Refs. IC aAd 2G) euFreAtly a¥ailahle te ealeulate pPehahilistie

~   seismie hazaFds feF sites east ef the Reek;y meuAtaiAs (EasterA YRited States).

11 Either ef these Metheds eaA he used te earry eut the fellew i Ag ealeulatieAs, 12 with the apprepriate set ef limits asseeiated **ith eaeh methed. 13 14 o~::m:1~1::1:o.w1::n.1)::1:,::::::a.ii:mijte¢.;tia.6.1:1:t::1-,au.:r1:)tA:ta.t1,m11:n.~:il)he.::js.s1niana:m:::t.o 15 1111:~::e1tr1.:111:~Jiffl~:S::::::b.n11g:::1~:1:i1ta!:!:P:1m:§1:11::;: ill!l~l!lil!!t!lil!:!t::11:~IJ.:[ltj:j::::I 16 liiiiiii!il:1:11:::1:1:1::::::111limiill:i:,:11:i:2tiii1:imi2:[l ilim::i1:n:::111:1i1:1t1::::::11,::imllRl]1iii1ii::~: 1 1 1 17 iiil~)iiliil!li.iin:,::IU.liJ:::::1n1.t~illiilnSii.];:::::::iti::::::1::1.:::a::1:::::::1ijb.]::1:111i:1m:1§::I1m11i:'ti.:n::::::11,::::::~t.i§l/g 1 1 18 1.9.nt.1:n1n111::t1~,;1:o.wt::::i1se.11:::::::a.tt1t::t.H~:::,:w~,:t.1,r1t::1::¥-:s:~::::m:::::::m1,1:t:i1.tt1::::::ss.1:tH1:jMti11,,::::11t:11:netb: 19 111:::::::,i1:11:::::::,1:,nlili1t::illili9.1:::::siii:I~ti;ilimlinil:::::::1111,1:in1.r:11::;::111::::::111111:111

• 0 1111r,::::::::,:ni:1111Jle:,:1::::::1::i11::::::1:r1,:1:[11i.11::1:11r.1::::::11111:*:1 21 22 111>>:::m1=::;,i:t:ttmtm1,n11rm:::::1:::::'1i!:i~1t:1111:!:it;1i~j1t1i1~:,1~::l:!1:i,::::::1u1r1m:::11:111:1:nn~mi11:!:ri.1:t@1 23 :11P:~l!!l tlillI~:~:fillt::IIII:l!:~gg:,,:!F:l~l:!i~l:i: wl:1::::::l:l!lttirn::1:1:!:@!!~ffl!I.~;

1 24 i§iilll[lliJllir!il!iil!IIliinIIiiiil:nilli:l:111111,111:i;n1:1w::11~1:::::::11n:::::::11:i::::::1;J[ill:I 2s 111::::awo.:ruiw 26 27 OOhi.l[ifo.i::liW:1:n1::::::r.ei'il.:1:i:::::::[iiiuJ.lt.:lb.i::::::1it!:fffi!1:nei:1.i;:li1:ri!I0:fllt.li 28 11i.,~1m.1:o.1::::::1a1::::::,:,1.1:::::i11,u:n1::::Ji.ml;H):~d:m1.1ijri:;.' 29 30 ;r:wt@@?tliital:nti,ta.tt.t:ilon[Jti.efi:@1roUfid:@MQt.:t:d.re:-.,$u.r.<<:r@PtP.v:tt:IJ.dJr::t:n 31 lillIIIIElliIH.:!:in::!:J:l.l:fllI[liitl:1:1:1:1:::Iinsl!l!inllllii'.iml:11r1iilit 32 C-2

NUMAAC Comments March 19, 1993 Line In/Line Out 1 ~??'??t?==a1111!t.esa,ui:i=1.z111=t:-.s:ti1:1:,i:turi:1fr:matct:x:io.1,:fa*,ni1:uaet-i 2 1,:~:ifil1ris1:::::=111:r1,:~:, 3 4 5 6 J;b.e:J~1::1re.s:u.1t:$;:ta:re.::uiJJ.d::tt1td.1:t1rmi:ne.::tthe,tss.11:a.na,tt.o.md.<<vel:bpr:tn1n:**1:$.mt~tdl1zara. 7 l!ili]iliWon:::=::1,1,:[ 8

-~  11 12 Step~,.                 Pe,fePM the site s~eeifie haza,e analysis Ysing the LLNL e, £PRI methee ane asseeiatee Etata. F,em this analysis, lili:[1:ii[llllllliirl rliilliiII!nllilii:l:l:!:llltl compute ili[l-R!lill@ median hazard curves for the average of the 5 and 10 Hz iffi'hibfabdH:z;.:S:tHI spectral *,el eei 13                         ~ ~~~*~,~r:~~,~~~~' s~ IHn~ !!!~tf~~u1J1iN~f~ppp~,~~~~;y. That is a 14                         ffil!il curve!: show-i-n§ the PP!!PP:i:~:1:1:tlffli:ln annual probability of 15                         exceeding various levels of the average of the 5 and 10 Hz i.RJI:t 16                         :1,1::::::1:f:~::::::11 spectra 1 tJel ee 1ty 1il1:1]~ril:!li!i.

17 18 Step ~I- Using the apprepriate annYal prebability ef eKeeeeanee letJel 19 liliriiillfliilil!iJ:::i11~:::11~, 11:111::Illii@lll1IJ::l:rie:::::111:1:1:::::1ui:1:11111 1 IIIJl,.O 11:!:"'=l: .Pr (e.g., fep the meei an s~ haza,e eYPtJe ee,i *,ea f,em the ~I LLNL methee, Pr is U 4 aeee,eing te Figype 8.3 ane Table 8.1 ef 22 A~penei K8), enter the hazard curveil of Step g a-t---Pr to determine 23 the Ill corresponding spectral acceleration:iirtJil,e11\liil:llI:lnf.ttt* 24 25 §lililllmltliiittll,iIIlfiiI=iiii!limI::i:lifllili!lli,il::lti!:l:lili,:imi@l:Iliiili:l:Jltilil!il:!lii 26 fiiii:!Ili1Mii1s.n:!:!:s.miiltillllii::!Il1:ivil:i::::::~iliiimlniltlJ.ln:l'llllliiI::1:i::::::::::::rn§illJ.liii-illlii:llll!lli!i 21 1,:1::UWiin:,:1:111:1of1l 28 29 ~?r==:::::::,:,:==::=nf:rac1H:aii@::con1.r11ut:J:tin:(01:@e1c1,:,ia1nii:t1111#:a1:st:ancaxpa:1:r:::i:£oi:1&.,i:Ss,1:i 30 31 1::i::=t:::::ii:=11,,IIm1.1:11:1111:11uI::1:1:111111:;1:::::::1n1 32 C-3

NUMARCComments March 19, 1993 Une ln/Une Out 1 2 3 QeaggFegate the median ef the a¥eFage ef the§ and IQ Hz hazard 4 eYr¥es as a fynetien ef magnitYde and distanee I):)' ealeYlating the 5 eentrihYtien te the hazard fer all ef the eaFth~Yakes in a 6 seleeted set ef magnitYde and distanee hins te determine the 7 Felati¥e eentFihYtien te the hazard, H., fer eaeh hin eenteFed at 8 magnitYde m and distanee d. M. is the annYal prehahility ef eMceeding ST~r) cempYted fer a hin at magnitYdc m and distanced. ~~ 11 11111:1:f;:Jil]J::I:MiU1it:IIII:ili:iiiiisillfiJl1ilI!iiilil:Ilii11:):1i::f1t::1111:::Iiiiiill:::::m,1:1::10 1 12 :1::,tt11::1::::::111,11:;,:111:rJ:1::1111:1:r1:it 111111:1i111i11,t:11i11i:11:,:1:::11,:11i!::111:1:11 11 1 1 1 13 :11::::::1ni' mtfiliJ. ;:::::11:in;]hi.iiffl1;;~::::IfR1Mii§II1itliHIMi!.l.l.iillHi.llliP.i§l1!iJHil 11 1 1 14 mis.ii~::1,11::::::111J:~Hliiliiiiflli:liillIIIil'!li].lin'rnlll1i.ililllJlnI!Iiiill::::!fliililltll!@i 15 l!:1t:1111:i:l:i:ri:ili:ilin1:ii~:ii¥.iitt~l@:i:ln11ilili1:il:ii:1i1til:IMi]j:g1jtil:':)i1sl 16 mi1.i:1::liPl[iiill[IJii:iii!2:,::::::1i:~it::11:,:mi11:1::::::i1:11:1:11r:1111:1:1::111,,1::11::1 17 iiiliiii~fii:::itst!lii:flililiilll 18 19 ~~ 22 23 Step 4. CempYte the magnitYde ef the eentrelling earth~Yake for the median 24 estimate Ysing the eentrihYtiens H. eempyted in Step 3.

~,

25 28 29 A E E mH./ E E M. md md 30 31 The distance ef the eentrolling earth~Yake from the site is 32 determined frem 33 34 j5 1ii=O E E log(d)H r/ E E M. 1111 37 Ill d C-4

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 2 3 4 5 6 7 l :ti:1: : 1 : : : 1 ::::::1111:::::::Illi 11 8 nne:rr:m:r:::=:r:n:n:n::rr:::n 9 10 11 il:iiii~: 12 ll1~:1::1:~!;ll 1' 1!M:~1-: 1 15 111::::::m1iiIIIJ::111111::::::1a:I:1:ijlt:i:::1:1::11IIillil]ggl@rmi:111::::::111t19:i:11:::::::111;: 1 16 17 °qrrJ,rb,P.1 18 19 jt'-i!i:,i:il/:::::it:n.:1:1:1i:::1:11.1::::::: 20 21 22 iltttJ= 23 24 RJHJ.:::::=m=,:=:=:dl:Cm:dt:l 25 26 27 28 29 1111::::::1:~::=:::::::::::::rr:r:111::::::1:11:1111111:1B1:i:111::::::111111:1:111:::::::1=:n::r111i:11*i ::*:111111i:1:11:1:1, 11 32 111:111:,1:fiiiiir::1:111:t:i]it:,1:::::=1111t:1ii::1Hs::ttliniilIIIIIlliilliii 33 irif!iiilJ::1:*ti[l§lfiisiilinii@:::::::::-:mn:i:tiiI;iliilllililml.inll:I::iit 34 35 36 37 38 39 40 =1:l=l:*i*l:*=ttl-:-:---a~t:m!tl!l:l9!l*:i!l!l:Hl9:l:i:11111ffi::::::1!:i::111:s::::-:11lt!S!itib:-:d1l:U!ll-t~! 41 *--- 1,11:1:r11]::1nsi:1111111::J1:i::::==111:1:11t:11111:::::J111::::::11m11d:e 42 1111:11 C-5

NUMARCComments March 19, 1993 Une ln/Une Out 1 1:mtl(=E~).:::tt:=rn,tiftactlo.rrwd.01tt:t:butt~nbtor:1tfie:::::,,0£1lr:::1ngand1iii=,111lc 2 fiJP.;JllWtiili.lt]ffl,HildQt.@:1.m,:Uiffir:61:te.ltiJ,J\~¢.Jliitt.f. 3 IIiiliIIIi f! 4 5 1:1,$0J:Jrmtnn,=m111111111t,i'.'.::&nt1:11Utl1ft~rm,inrr.tiw1.i.il1n11.1.*nwt1se:1-:11 6 ~itt<<tJliil.\\\/,\IY@8i'tlffi[l¢\f.ejEt.¥ifd.irtml 7 8 11 12 Ste, 5. Ysing the same P~ and ste,s I thFough 4 as aho¥e, also deteFmine 13 contFolling eaFthquakes for median s,ectFal Fes,onse for the 14 a¥erage of the I and 2.5 Hz s,ectFal res,onses, and foF the median 15 estimates of the ,eak gFound acceleFation. 16 17 Ste, 6. The ground motions coFFes,onding to the controlling eaFthquakes 18 are determined as outlined in Section 2.5.2 (Suhsection 2.5.2.6) 19 of the Standard Re¥iew Plan. ~~ 1111::::::z:;::::Illl)ltfl!tn@::tlSIJl:rlipifi.~i:t11n11,11:,,:::q1n::n1tlliitlllfiil::::::rqi,J)nlI:lfiitl1rlii.l)ijii. 22 liii!iillil!l!!:ti,tliUna:~:1t:1t::11:::::::in1::1]lli.il!iliri!flil!lliilillit:iiiiltili 23 iriill@iIIillil:liliili~:9:1,::::::11:r1sit:1::::::iiliiiiii!ilI::::,:,1s1iuil:/!!lililti:::::1,::::::1::::::iiiJll1:,: 24 iiisi,IE!isi:i\i:11:11,111:IIiililnil:iliiiiPiiii:thili:i:i\iisiiiiliiiliiiI;;IIIIIiltiHil 25 ffl9;:i:11liIIl ilirlll1il[]ln:filiilllJli 26 27 G.2.2 WesteFn Y.S. Sites 28 29 FoF the Western Y.S. sites, a JJFOhahilistic data hase, sueh as com,iled 30 in the LLNL OF EPRI studies, is not a¥ailahle. In a Fegion of acti¥e tee 31 tonics, theFe is less unceFtainty ahout the significant contFihutoFs to the 32 seismic hazaPd, and the controlling earthquakes can geneFally he defined C-6

NUMARC Comments March 19, 1993 Une ln/Une Out 1 detePmiRistieally. FeP PegieRs ef leweP, less aeti¥e teeteRies, aR aRalysis 2 s i mil aP te the eRe eYtli Red abe*.ie i R Steps 1 4 eaR be pepfePmed.. Step 1 weYl d 3 be emitted aRd the &y-1,. le¥el Ysed weYld eePPespeRd te the ¥a1Ye seleeted fep 4 the SSE:. 5 6 C.3 EXAMPLE FOR EASTERN U.S. SITE 7 8 To illustrate the above procedure, calculations are shown here for an eastern U.S. site using the H:Nt ill:{: methodology given in (Ref. IC). ~~ 11 Step 1 is emitted. 12 13 lliet:J@Ii:illilII:WliIIIIIRllil!t~::imi:iii:::::111111::::::m111!l1Jinfilil1::::::1111::111::::::111!mi:n1f!I! 1 1 14 tiii:itlf:;i,ifiiiil:J!ii:I:]iq::i;::111:i111:~:::::::::111i:11:1i11mi::::::111::::::url11:i:lii 15 ii:iii:irill:i:iiill::inii:1.:1ni:11:::::::1:1:i:::::1:1:1:::::::1i:1::1111:::I:I.IIll::~rr:im111::ri1il:1:1:t!!:ii 16 liiiirii:i1ilfll:iii:Iiiini!:lllimllliliiiiI::1:!:n:i:f 17 18 Step 2. Table C.1 gives the annual probability of exceeding various levels 19 of the average of the 5 and 10 Hz spectral ¥el eeity :!!!@J:i!!t~:!:9! hazard curves from the H:Nt :1111:i study. ~~ C-7

NUMARC Comments March 19, 1993 Une ln/Une Out 1 Tab] e C.1 2 3 Average of 5 and 10 Hz &., I, Curves for the Site

• 4 5

6 Annual Probabil i ty of Exceedance 7 (Median) 8 9 C12 13 S.teplfflii:i:tit~iln1jt,1:n1:@ljll:1.@:1¥:J.:ijit.ff.j;;l.fft£l1f.,r1nst@Jt=rto.tiuJ.ml:~;t:;WJl~ifavil:MiE@i 14 1:~:lilll~~fl11!I1: *J1nterpo 1at i ng!ii the annual pt=ebabil i ty ef e>Eceef:lance 15 fPr) ¥alues gi¥en in Table 8 . 1 frem Table G.l, the corresponding 16 va1ue for S.,fPrt l,,.:Ufflli is as ghen in Table G. 2 111:I:ii~i~i§. 17 18 19 20 Table G.2 21 tt: 24 Median 8 25 26 Step ~. FeF this e>Eample, te f:leaggregate the hazarf:I anf:I f:letermine the Hi..rT 27 it is first necessal":)' te cempute the centribution te the average 28 hazarf:I fer the § anf:I 10 Hz spectral ¥elocities l11J.:~i:;:f:~!!:99.£fli.ill 29 iiill::iI:iiiiiii.:;:;::iiiimii[:ii1:iiiil:i11t:i:i:i:~:i:;:::iliiii::1:1:Iii1E::~iiiiilii 30 iillti:111:1:111:11t1iiltlliii:1111111::111nsiIIiiJ:1::::::1,:ttffilltiililil:t~::1 31 lilim:lilffl1Ji[IiiruhiII11:~::1111::1IliliiltM::;:111111lri1iliilI[ for the mat r i X Of 32 magn i tudes and distance bins such as given in Table c.a.J. 33 34 35 Table C.31 C-8

NUMARCComments March 19, 1993 Line ln/Une Out 1 lt~P:l!l"=!:il Magnitudes and Distance Bins Ysed iA EMaM~le 2 3 Distance Magn itude Range of Bin 4 Range of 5 Bin (km) 5 - 5. 5 5. 5 - 6 6 - 6.5 6.5 - 7 7 - 7.5 >7.5 6 0-25 7 25-50 8 50-100 11 100-150 150-200

           >200 12 13 14           For each bin, a complete lnilii;fJ::1 :11l1 hazard analysis               1  1
                                                                                                                                    *  !!I performed 15 to give the contribution to the hazard from all earthquakes within the bin, 16 e.g., all earthquakes with magnitudes 6 to 6.5 and distance 25 to 50 km from 17 the site. The results for this bin are given in Table C.43.                                                             .;.;.:

18 19 20 Table c.+J. ~ 24 25 GeAtFi t:)yti 8A te the IA1.nll$g~:~m~:s Hazard From a11 Earthquakes in the Range of 6 s Ms 6. 5 arid .. dfsfari'c'e*s 25 s d s 50 to the aver a e of the 5 and 10 Hz Spectral \teleeity Bsll:<<ri!:,,::g{ 26 Spectral Annual HediaA lfeian 27 ........................... <'..! .......ST.... .

                          \teleeity                                                                             Probab i l i ty o'f".,.,.

28 AijeeJirijtmonits" Exceedance 29 30 31 32 33 34 The value of M., 'L (~~~P. annual probability of exceeding STfP~t l!F 35 li.!(1111) fffl this bin is obtained from Table C.4;1 using the STfP~t f,1(!111! 36 values gi¥eA iA Tahle G.2 and computing M., L ......... *.*.* by interpolation. The value~ C-9

NUMARCCommems March 19, 1993 Une ln/Une Out 1 for w_ ii@ for this bin ape l:i given in Table c.51. 2 3 Table c.q 4 5 Value for M. L for the Bin 6 :S RF;illil _:S 6.5 and 25 :S d ~--* 50 for the Example S1 te 6 7 8 Annual Probab Ui~Y of Exceeding s.-{-P,:t 9 S~WBPJ For a Bin 10 li.QE 6 5.l0E-6

• 13 14 15

   -s+te.

Tahle G.6 gi¥es the complete matFix ef the Hmir ¥a1Yes feF the example 16 Tahle C.6 17 18 HIIM ValYes feF All Bins Based en the Median HazaFd 19 (Nate: If Mm s l.E IQ, it is listed as Q.) 20 21 Distance u~--~+

                                                         ,,i~

n~-~~

                                                                *-  ::1-
                                                                          -£
                                                                          -   -o~M" 22   Range Bin
                    §     §.§     6.§      6       6      6. Ii        6.§       7    7 7.§        ~
         ~           2.QE   §      I.IE §          2. 4e 6                  G            G           G
         ~           6.2E 6        8.Qe 6          §.Qe 6                6.§e g          G           G 25      §Q lQQ       6.QE 7        2.3E 6          6.8E 6                8.4E 7          G           G 26     lQQ l§Q       I.SE g        l.6E 7           L§e 6                2.8E 6          G           G 27      l§Q 2QQ           G         I.IE g          2. IE 8               4.6E 7          G           G 28        ~               G.            G                G              6.QE 9            G           G 29 30 H ----1*11111-11111:~JE C-10

NUMARCComments March 19, 1993 Une ln/Une Out 1 2 !IIJ~1:: :~=i:I 3 4 5 iitii:-IIW;t.o.iiive.iiiJjlii:liillllJll§ill111F"i1Eoiie.iiina1:jg 6 7 8 9 12 13 14 15 16 17 Step 4). To compute M, D for the example site, the values of~ llrnl!l;l: 18 given in Table C.~I are used with m and d values corresponding to 19 the midpoint of the magnitude of the bin (5.25, 5. 75, 6. 25, 6.75, 20 7.25, 7.75) and centroid of the ring area (16 . 7, 38 . 9, 77 .8, 126, -~ 23 24 176 and somewhat arbitrarily 600 Ill km) . Thus for the example site, the eeAtrelliAg eart~quakes, iA M, D values, are given in Table c.-71. C- 11

NUMAACComments March 19, 1993 Une In/Line Out 1 Table C.1§ 2 3 Magnitude and Dist~J:1£~ of Controlling Earthquake From the 4 Y:N'= BRRI Probabil i stic Analysis 5 6 Based on He~iaA Miiri Hazard Estimat'ei 7 M :1 I 8 -~ 9 D 12 13 14 15 IH:::Isll!nj:fl1!:it1i::Ialfl!Pltil!:t1m!isII:1,1P.m!::::(1gil!lilm!lniliill!ll:1::1:1.1[lil 1 16 @lilffljij¢11lf.1Itlh~ftl.S.l::JJ:=,:11,t1.rmito.ld:J:f:rpm:t1Htff:d.il.Ufflllt:i:1rutrnby);~111m1@::t~p9.r11¥ 1 17 AfitJ.:b:1:r1:;11iri@t:::::rn1bJ:~fil!tl::::n:m111:11.11:=:::::e:t:i:}@}lfi~'f'l:(U)Ml9lt(ll.ifi):)::;e.sm@t:lt*II:lt:P::r1 1s l:;,:IIJl!I!lffl:t11!:lmi:S:H1IM!l!!'itlil!JiJ:J1!111@::tlfi!::Il:rMIBUil.UIJ1V:ll!!iUml1.J!II:J!lt:lfilliffl!il 19 1r1l1i!:J1:i:!!::;ii,~:::::g*=11n1.n1i::::::l9t::::::~l!::::::1M~ri1.1@t:1iie1111l:::::=:j,g;jJ:1111\1'1n:::::::~§:J1i:i:in.1:'~: C- 12

NUMARCComments March 19, 1993 Une ln/Une Out 1 ffia&1:wr1=i:z 2 3 4 5 FrJ<<l:tonl.lll~;J;;aJlllliJil;l;!llllll':!!t.*~*:::llllllll:1:i:§.aqrce. 6 7 8 .0 9 11 12 13 14 C-13

NUMARC Comments March 19, 1993 Une ln/Une Out 1 2 m,~~:i?~::~:~ 11~\\lll11llllll1lllll11\11111ij1ii!!,\ll::11 3 4 5 6 7 8 -~9 12 13 14 1§1

1:1!1 11:1 11::itli 1:1::111 1m~:1:1 15 1111 1na:1z 16 17 18 Iii! llf!P!il 19 R 11:11:1 20 21 s

                                   *:-:-:                                                                 11:1:11 22 23 24 1m:,:                                                                      11:11:1 25                             :~:1:11                                                                    1:1119

~6 27 28 29 30 31 32 33 34 35 lniIIlll]liiiiiilI!::nliilii:i:1::::::1:iiil:11m::::::i::1::::::s1liliiHltiil!:ti,11::1:11::11::11:;::111:111,:1:1i:s 1 36 !P!Slt!lil!li:11:t!!:!~t:1e1,ui1s::11~till!If!!ll=!t!II!ll!!IIIIII!!!!ll!:r!!:1:11:::Iltll:IIIl1IIII:! 37 11::::::,11r::1i::111:::::::1 1i11::~::ttim1::=::1t11111:,:,::1,1111:r111:::t11a:r1,n1:11::::::1;1,1u:111r:,i::::::1:1, 1 1

~

38 iiiiUiviiliiiil::~:nil[::i:n:Iliiiiliiit::1:f:: 39 40 GeRtrelliRg earthqYakes fer ether freqYeRcy raRges (Step§ ef G.2.1) are C-14

NUMAACComments March 19, 1993 Line In/Line Out 1 ealeulated hy repeatiRg the ahe¥e feur steps. 2 3 C.4 EXAMPLES FOR WESTERN U.S. SITES 4 5 m;::::1111im.J.:Hil.llin::::::11::::::1.li.Iil'=l:iiliill'::::,11::i11m11:,:1:fllii.iltl:!:ililiil:J:11.::Illli::11t, 1 6 11:1:tf:$1::i11,im:::1.Het::n1:1,m:t1fa,1:~mt:1:11:tic*rrr1:,ato.uti*111iuit111:i1n1ce.u.t:ia11<<r:1:l.it:1,n 7 11,1:111u::1:i1:i:tI]l!II!tHli!il1m11:1:rn,t1:111:1::11!=1:t:tilliil!III1niiiffll!lllmllt:1n 1 8 illlllit:i:1mr~::1:11:~1l!M11:~]:::,i:1!11i1li!l@limt.llillHli]l!:111nsiI=§i~P:!=ll1b.11i:1t:§1t:91:11m -~ 11 12 SiRee a geReral appreach fer the WesterR Y.S. sites is Rat a¥ailahle, twe specifie cases illustratiRg determiRatieR ef ceRtrelliRg earthquakes are discussed helew. 13 14 C.4.1 Diahle CaRyeR 15 16 The Diahle CaRyeR site is lecated eR the CaliferRia ceast. A legic tree 17 appreach has heeR used te assigR weights ta ¥ariahles asseciated with faults 18 Rear the site aRd determiRe maximum magRitude distrihutieRs (Ref. 3C). The 19 legie tree appreach was alse part ef the ~rehahilistic seismic hazard aRal ysis. The result was that the Hesgri fault zeRe was the mast sigRificaRt seuree. The eeRtrelliRg earthquake fer the Diahle CaRyeR site is a magRitude 22 7.2 e¥eAt eR the Hesgri fault zeAe at the clesest distaAce ef this fault zeAe 23 ta the site (4.§ km). The eeAtrelliAg earthquake magAitude is larger thaA the 24 maximum histerieal earthquake (the 1927 magRitude 7.Q Lempee earthquake), 25 whieh may ha¥e eceurred eA a structure related ta the Hesgri. 26 27 C.4.2 WNP 3 28 29 The WNP 3 site is lecated iR westerR WashiRgteR aRd lies ahe¥e the 30 Caseadia suhductieR zeRe. The staff eeRsidered feur eeRtrelliRg earthquakes 31 fer the site (Ref. 4C): 32 C-15

NUMAACCommenta March 19, 1993 Une ln/Une Out 1 1. The applicant prepesed that a maximwm Pandem earthqwake in the crwst near 2 the site is magnitwde 5 1/2 te i. This ea,thqwake is hased_en the largest 3 histe,ical earthqwakes in the Geastal Plain seismetectenic ,,evince (ahewt 4 magnitwde 5) and the peselwtien ef geelegical stwdies in the site regien. 5 6 2. The maximwm earthqwake asseciated with the 9l;YR1pia Lineament, which is 7 35 km nertheast ef the site, is a magnitwde 7.§ based en estimated maximwm 8 rwptwre length. .09 11

3. The maximwm magnitwde earthqwake fer the intraslab swhdwctien zene sewrce is abewt magnitwde 7 1/2, based en the maximwm histerical event asseciated 12 with the Gascadia swbdwctien zene intraslab sewrce (the 1949 magnitwde 7.1 13 Pwget Sewnd earthqwake) and cemparisens with intraslab sewrces in ether 14 swbdwctien zenes werldwide.

15 16 4. The interface swbductien zene sewrce is capable ef great (larger than 17 magnitude 8) earthquakes. This maximum magnitwde is still wnder revie~~ in 18 light ef engeing geelegical stwdies. At this time, the staff censiders 19 the maximwm magnitwde te be 8 1/4, based en argwments abewt the likely -~ dimensiens ef rwptwre and cemparisens with ether swbdwctien zenes with slew cen¥ergence rates. C- 16

NUMAFC Comments March 19, 1993 Une ln/Une Out 1 REFERENCES 2 3 IC. 9. L. BernreYter et al., "Seismie Mazard Charaeterization of &9 NYelear 4 Plant Sites East of the Aoeky MoYntains," NYAEG/CA &2&9, Janyary 1989. 5 6 ~it. Electric Power Research Institute, "Probabilistic Seismic Hazard 7 Evaluations at Nuclear Power Plant Sites in the Central and Eastern 8 United States: Resolution of the Charleston Earthquake Issue," NP-9 6395-D, 1989 . -0 11 3C. "Safety EvalYation Report Related to the Operation of 9iahlo Canyon 12 NYelear Power Plant, Ynits 1 and 2," NYREG 967§, SYpplement No. 34, 13 Jyne 1991. 14 15 4G. Letter from Marvin Mendonea, NRG, to 9.W. MazYr, Washington PYhlie 16 Power SYpply System, "NRG Re*1i ew of Sei smi e Report for WNP 3," JanYary 17 4, 1991.* 18 19 20 *Availahle for inspeetion er eopying fer a fee in the NAG PYhlie Deeyment Room, 21 2129 L Street NW., Washington, DC. C- 17

NUMARCComments March 19, 1993 Une In/Un* Out 1 APPENDIX D 2 GEOLOGICAL, SEISMOLOGICAL. AND GEOPHYSICAL INVESTIGATIONS TO 3 CHARACTERIZE SEISMIC SOURCES 4 5 6 D.l INTRODUCTION 7 8 Seismic sources are ijgfj!!i,IJiilUtii!i!!i areas where future earthquakes are 9 1 i kely lJ,P,lij:1,,a to occur tditffl::1ijjjr5jj;ijj),igijij;,r,ri<i1:tn11~:$.:::::::1,1:tm.jxlrim 10 mi1.n:1::t@II* Geolog i cal lIIliiilli~:111::~: and seismological investigations provide e 1 the i nformat i on needed to :1:sinllil!tii!il characterize source parameters, 12 including the size and geometry of the seismi c sources, earthquake recurrence 13 models ~iii:!:, and deteFmi ni sti C SOUFCe eaFthEtuakes (DSE) lffl.iJ[ffl9!1:::::ffll9i9,jji,yjl,~. 14 The amount of data available about earthquakes and their causative sources 15 varies substantial ly between the Western U.S. and the stable cont i nental 16 reg i on (SCR) :1:"1:iJ1::::::11:1::::1~i::::::11111=]ll~i.i:t:1:1:n:,:1@ and also fF~m Fegien to Fegien 17 wi th i n these tiJFO ad aFe as . lsrll,ifffliii:i:I:11:n1:11::::::1r1::::::11:i:!:i1:,::e,:~;;::::1:1:11:n,:::::::mu~:n1::::::1I1 18 llfmlf!Jml![:j:[li!ii!ti,JjJp!j!gj]!!liijjfji,jjjjp{tl]i! In act i ve tecton i c reg i ons , the focus 19 will be on the identificat i on of oot-h capable tectonic sou,ces and seismogenic 20 souFces :l@i[Jl'!, and the methods described in Sect i on D. 2 of this appendix can 21 be applied. In the SCR east of the Rocky Mountains, seismogenic iliil:1,:;f.\i;ilil e2 ~ij~[!f!Ji~ sources JJl ay a significant Fol e Jjjijj:j::~:@ii;\j\J!S~f:J)i because of the di f fi-23 culty in uneEtuivecally correlating earthquake activity with known tectonic 24 st rue t ure s iJfflllilll:l::111.:::::enliliii:11i:I:::1nill:ilii:ti§8ilii:i!Bl1iilifii:iiiiiii* 25 26 In the SCR][ a numheF ef significant tectonic st ructures exi stji::::I1nl!Ilh1ili 27 -t-ft-a-t. have been suggested ~:n1itP!l~I~ as potential sei smegen i C SOUFCes ~iiill:i 28 l;,pjJ.;~I (e .g. , the New Madrid fault zone, Nemaha Ridge, iii Meers fault, AamatJe 29 fault 2:ene, Gl a,endon Linden fault)

• lfil.ii[:jrnlii);!ti!i:;jlfflijjjtlffl.[!;/j::niiiiri.iitnrn::11 30 !fililO11.M$.!li¢1!:ta:1:1:11:J::1m:i:1:t:111uttfilmt:t:There is no ~ 1:u:111:1 procedure to 31 follow to characterize the GSE !!!lj.:mffln magn i tude associated with such pessible 32 seismegenie sinliiiili:[i[i:ililm~[s sources ; therefore, it is most likely that the 33 determinat i on ef the se i smegen i e natuFe of t he source ime,:il:i:ii will be D-1

NUMARC C,ommerrts March 19, 1993 Line ln/Une Out 1 inferred rather than demonstrated by strong correlations with seismicity 2 and/or geologic data. Furthermere llt!!YI!, it is not known w~at 3 relationsW,:J/i'- exist between observed tectonic structures in a §+'Yetl 4 sei smegeAi e iii:1iimil¢. source IJ::t.:6.i1iniltfillil¢.:~ and the current earthquake activity 5 l easel y eerrel atee:i tbat:rmayt:ben,assnei::atffll:t:wi th that source. Genera 11 y, the 6 observed tectonic structure resulted from ancient tectonic forces that are no 7 1anger present' -3M thus the Stl'uetural 1::::::11:iisliiri:~::1: extent may not be a very 8 meaningful indicator of the size of future earthquakes in the source. Careful -~ 11 12 analysis of the historical record and the results of regional and site studies and judgment imlliif:i play key roles. If' on the other hand, ~ iir.i strong correlations and/or data exist between seismicity and seismic sources, then approaches used for mill active tectonic regions can be applied. leiill:iilil!:! 13 liRiiilili[i1iii:i:iii!inii:i:::1gli~I:::::11,::i1::1::,:1:iii1lIIliliitiliiilii]li]112:1:9;i!ii1m~::::::::1;:1::1m2l:eiiilsilUiliil 14 iiiitilillsl!If}:!1.ii!1iiii!s.iil!iiP!nitiiiii!iflilM1ii)!'i1::::::11:::::::11Miiili11:1:::J,:1:1e1eislilillllillnii 15 :1§iliiiiiiIIlililiii]I1lill:JiiieiJ:imiiliiifiiilliiilnW.))ilnf:91}jj:~:1,:::::::1:11:11::11:1i:iJIEtniIIriiiliilii 16 i!I:lliii\ililii@i!liiilliiliiililiiil:Jl:J}lili:iI:iiilIIililJliiiiiiillilillliiJilf:illiii1iiJliilifiilil:1;,:;::r; 17 i1iiiliiilill:liliiiiliiiiliil!iliiliflli'i'ilililllilinifilji:ijf,91jiiiliiiliiilil!iliiililiniIIliiiiii~i:t11:1 18 ii.iiiiiilililllliiliil1m:1.:1:::::::111:111:::::::1111ij::1:1]lll]Ifi[ilihi[iiiii:Illiimi:i:lliiiililiiljllimiill!iliiililsiiPliiYiiilsi 19  ::un:1:ar:r1:n@fJ:!IWflilil::111:111ii1:i:111:,1::::::1i:11::1111rt:::1~:~::1:tJ:tS.I=t!&lfiifl!:i:!:!!@l:l!i!R!!!lJ: -~ 22 23 UiM#.i.l:qiji.lijl=:::::gfttliW'fil1:1m1::1:J:i§U.r4tjfii::~:1ma1:::n1au.:utem:::::::111n::m:1:1:1g1J,UtJlti¥J=$l:R 1n1:11::],,Jl:111r,:~::1:::11::1:11J::r:1:11111:11:1:]lliilJffl!l!ll:efft:l:111::::::1111r1:1:1::::::1:1:]IPllli:S!IIt!¥IIIH!

1:1:~g::tUt!!:i~:i:s.1.l:i::~un,r::1nrJ:fi~P:!iti:!iU'i!:~:t:=i:n::rlb.!t:!1:111m11:~:tlM@::::1:j::~!blP!S:1::nte:::m:11:1::1!M:P

    $.1ur¢t.$.':tamu::agt~rm:1:rtm1:~,::::m~:~:u:~::~:::::::'=n:R:~:p:,:1~+~~]:~1:~~r=roy.pf:,q#:~t:::,:,:+'-:~:::u~:n::::mf-g'-:::::::~¥:g1::m::P'-W 24  H!llffll!:!1::ett:::::1:tl:Jiitlnlliil!lllf=llllll!E!ililiI:iliHnl!lil!?.!fiJi!ilii1!jln.!f:lllllOilIIl!grsl 25  1::111:1,11111:i111@

26 27 The following is a general list of characteristics to be determined for 28 a se i sm i c s OU rce imilnIIl:ilili!I!Illiiiii!ii!i:i!illliirfii)):[nlimiiillilii!iniIIiiiI}loo!lililil : 29 30

• Source zone geometry (location and extent, both surface and subsurface) .

31 32

• Description of Quaternary (last 2 million years) displacements (sense of D-2

NU~Comments March 19, 1993 Line ln/Une Out 1 slip on the fault, fault length and width, age of displacements, esti-2 mated displacements per event, estimated magnitudes per offset, rupture 3 length and area, and displacement history eF uplift Fates ef seismegenie 4 feles) . 5 6

• Historical and instrumental seismicity associated with each source.

7 8

• Paleoseismicity.

9 -0 11

• Relationship of -the faul di.l to other potential seismic sources in the region .

12 13

• 14 15
• ReeuFFenee meeel (flrequency of earthquake occurrence versus magnitude).

16 11:111:tliilir~::::::11,1::::::11::iJ:::rt1:]iiiilliililMJliiit@Jlillfilitltilll.i§l:!i9.J!sI:::111:1:,:11. 17 li::I:!i:r1:111:~Ililiif 1 18 19 .. Effects of human activities such as withdrawal of fluid from or addition -~ 22 23 of fluid to the subsurface, extraction of minerals, or the effects of dams or reservoirs. Volcanism. Volcanic hazard is not addressed in this regulatory guide. 24 It will be considered on a case-by-case basis in regions where this 25 hazard exists. 26 27 .. Other factors that can contribute to characterization of seismic sources 28 such as strike and dip of tectonic structures, orientations of regional 29 and tectonic stresses, fault segmentation (both along strike and 30 downdip). D-3

NUMARCC,omments March 19, 1993 Line In/Line Out 1 D.2. INVESTIGATIONS TO GMARAGTfRIZf IVALIATE SEISMIC SOURCES 2 3 D. 2.1 General 4 5 Investigations of the site and region around the site are necessary to 6 identify both se i SIRegeA i e iilllH!@ sources and 1:1:101:~te!n:I cap ab 1e teeteA i e 7 se~!Fees lt!Jiti!lil and to determine their potential for generating earthquakes 8 and causing surface deformation. If it is determined that surface deformation 9 need not be taken into account il]:~llli§:;:~i, sufficient data to clearly justify eo the determination should be presented in the license application or early site 11 rev i ew . liiilil:l:x:;Jiiilltlliiiii:!:1::::::111trmilllintiil]1lllif!linl:i:~:1t::1:Mi:li1iIIil:ti.lHiIIl:I 12 l.m:ifl !IJ:11::1:::::::11i:1:::::::1:1:11J1:1:1:!:::Jli.il1:11::::::11iiii:ii:t:111mlnil:i:litI1illlil:1m11::11i::~::;; 1 13 :~J;Qn;j:t;~:~lf9.P.~W' 14 15 IA the sitiAg ef AueleaF peweF plaAts, elngineering solutions are gen-16 erally available to mitigate the potential vibratory effect of earthquakes 17 through design. However, 5tie-h solutions cannot always be demonstrated as 18 being adequate for mitigation of the effects of permanent ground displacement 19 phenomena such as surface faulting or folding, subsidence, ii ground collapse, -1 20 22 eF fault eFeep . For this reason, it is prudent to select an alternative site when the potential for permanent ground displacement exists at the site (Ref.

   -l-1D). In most of the EasteFA YAited States SC.I, tectonic structures at 23  seismogenic depths, as determined from earthquake hypocenters, apparently 24 iilii bear no relationship to geologic structures exposed at the ground 25  surface. liil:11:1 11m1:1 ¥1oung fault; iiiilliili either do not extend to the 1

26 ground surface or there is insufficient geologic material of the appropriate 27 age available to date the faults . Seis11egeAie lliii§Jlli faults are not always 28 exposed at ground surface in the Western United States as demonstrated by the 29 buried ( hli AEI) FetJeFse seuFees liijJ.l~ll of the 1983 Coa 1i nga, 1988 Whittier 30 Narrows, and 1989 Loma Prieta earthquakes. These factors emphasize the need 31 to not only conduct thorough investigations at the ground surface but also to 32 identify structures at seismogenic depths. D-4

NUMARCComments March 19, 1993 Une ln/Une Out 1 The level of detail for investigat i ons should be governed by l/o.pj!]1 1Jgg 2 II the current and late Quaternary tectonic regime and the geol.ogical 3 complexity of the s i te and region. iliI{!lfiiill:J:9.li!1:sn!l!9!Yll!I]it;§1:11Uti.i!! 4 nt,lv.¢.inv:tjtt!li~lfftlj$.:JJ19liliYl1tJ1itlgtA1:l:il.i~i:,lri:f:tr@jlii:li.t?a$!Wlbtxm1rrt.~P.ilHi(t.4ittti.1. 5 lllli:lilJifl/ll!lllllilf!:ill!Jill!ll!!}:!,!,1]1~11~1E(ll:iil1lIIl!9II!ffill!RJI!IIl!f::f 1 Whenever f au l ts 6 or other structures are encountered at a site (including lt1:BI in the SCR) 7 either i n outcrop or excavat i ons, it is necessary to perform many of the 8 1nvest i gat ions described below to c:ie1R0Astioate l.!ll!!l~J!pl, whether or not they -~ 11 12 are capable tectoAiC SOl:ffCes 11,mli-lhifiisiil/lellllniliml/ii!ill§ni:IlinIIi:ti/t/li:/!Ilfiiil:ili:Ul:Il/l/il/llU::i/tiimit/1:lill ii:1111:1,::1:::::::111.:1:en1:imi:1:i1:~t::::1:~::,:;:1:i:11::::::1il/::Iliili/:/::1r1,;:f: Reg i on a1 i nvest i gat i on s extend 13 to a distance of m 20.0 km  :.:,:.:,:,:,:-:-:-* cm l.1!$ mi) from the site, and data -af1e i.hhU1'11Ilb.i 14 presented at a seal e of 1: 500' 000 or sma 11 er . ffifiiti!tf:iill!@l]M::11i.ii:lls.:1.1:u!ii 15 iHiill:/:/:ilij:Jpminnil:ttiIIl!lin!li/f.l:I/iij/:im!il]!iiiiiiiii]iiil/I:ll/i/inl/:§i/:Il!iiI/Iliili/iniri 1 16 l;iii.l.iUl.l]ffii§;jlffli::~:::::::::ttb.~::::a:1v,ii:1:s11:,::,u.:s.::::'i[i.biij)!;l.'t1::n11::111::::;:,::::i,11,r.,o.iniii1::v,::t1::J1'liriii.lii 1 17 itlii:IJI*::::::;illilisl;;99;j111l:Jiiiiinii/ilii!iiiit::::::11111:1::::;l1:i:i;iinliiig:::Illl;liil;llil;:::11111:1:11 1 1 1 18 11inliili.!B./I/:il!i:::i!IAl@il:~::,1::1:1::::1,iiilliiifilni11!11iiiliiiiil[ffini:iiiiiinlii/l!::m,i::I1i 1 1 1 19 /i!J!inf:!lslinl::li::illrI/Ililil:i:nil/ni:/1/i:!li:i/:!IIIIMittlis,nn:i:~:1:t:inii/:i:/1!!lil1J! tf;n vest i gat i on s * -~ 22 23 gioeateF c:ietail aioe l;,!~§i;!/!,i;::/ip£ conducted [~/n::::::Jl@l/!i~i !@/i/!/Jlj!/g ~i!U:!¥ to a di stance of 4G l;i km (~ Iii mi ) from the si te and ~ data presented at a seal e of 1:50,000 or smaller . Detailed investigations -af1e iliilil:/:!Jil! carried out j;/nlliifli IJ:11:::::lnli within a radius of 8 km (5 mi) from the site and 11111 data are ' 1 24 presented at a seal e of 1: 5000 or sma 11 er* ffinil;liliil.l:iillmi/tixii!j]iilJ/liilii!Mii:llll 25 11:tlIIl:!:tl:1:1;11n::1111111:11J:J]!!lil:]l!!iiliiJ6Jl9l)fil.:!:m11:1a:r111::u11111f:i:im:r11:n 1 26 l!!IP!lillliltimlft:ittliilliit:111111::::111.:tiriiM:!I!lirl:l!l@itff[IIJ{Jli!@illll!li!lP!i!::l:J!!liill: 27 p,e.:::ttfie:111.,11e1.:nt10.+e.fi1rJ.1t.1rl:z@:n:,:em$.li:J:f.tJ1:1,r.1.1.11::1@$.lll:tRJAn,:n1nlt:tHn1.1:r1:t111 28 are.t:(1ppto.prJ:ate:n~o.ttd;!het:l.e.¢.t.on:tp.;:rre.g}:meM Oat a fFOIR i A'le st i gat i OA s wi th i A t he 29 site aFea (appiooxiMately I squaioe kilo1ReteF) aFe pFeseAtec:1 at a scale of 30 ltSOQ Of SIRalleF. The aFeas of §,J:1~::pl:~:s~::n~iJ.MiHinv::::::1:1:1!t:1r11 investigations may 31 be asymmetri cal and seiii!li larger inii than those described above in regions 32 of late Quaternary activity or lil'Sfill!!re!li historical seismic act i vity (felt or D-5

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 instrumentally recorded data) or where a site is located near a capable 2 teetenie SOl:lfee s1:1eh as a fa1:11t zone r.ay:ut. Dat.1:,:::::1m.:n1:~v11~1:ga1tt1n1.r:wtfti:i:n 3 1Hew&s1111at1aJ::1:1,1roxur11i,1,::ru::::,,1:1t1w1n:"e.terJ.tfaneMt1ou:i:a:n1ewpt.e.s@nteau1t 4 1.mlt:il:1:i:111:t;tlllI:1t=Hil!ll::l:1m:~: 5 6 Regional and s1te 1Jnformat 1on 111:::1:111:1v1il:ii1I::1:1i1m1m11111:pyi;l'.1[1iilil:1:11 7 needed to assess the integrity of the site with ,espeet to potential g,01:1nd 8 motions and s1:1,face defoFmation ca1:1sed h;Y capahle tectonic so1:1,ces 1,:li, -~ 11 12 lililllil[ililf:iiil:ililililii:l:sn::::::,1n~ilitl inc1udej: determination off:t ( 1) the ieiitnrJl:i.ii 1ithologic, stratigraphic, geomorphic, hydrologic, geotechnical, and structural geologic characteristics of the site and the a,ea s1:1FF01:1nding the site, including its 1,u::m!!i:~::~ti!:inl geologic history, (2) geologic 13 evidence of fault offset or other distortion such as folding at or near the 14 ground surface at 8F near i~!l~!ln the 5'.li te lni1]:!:~1::::::g:::::111~:g;i,:~:, and (3) whether 15 or not any faults or other tectonic structures, any part of which are within a 16 radius of 8 km (5 mi), are capable tectonic s01:1,ces. This information will be 17 used to evaluate tectonic structures underlying the site, whether buried or 18 expressed at the surface, with regard to their potential for generating earth-19 quakes and for causing surface deformation at or near the site. The evalua-tion is to consider the possible effects caused by human activities such as withdrawal of fluid from or addition of fluid to the subsurface, extraction of 22 minerals, or the loading effects of dams or reservoirs. 23 24 D.2.2 Reconnaissance Investigations, Literature Review, and Other Sources of 25 Preliminary Information 26 27 Site and fllegi ona l 1::1:t.irit!U.rit::,,a:r111.a,n1:i)lili.iHPiilJ:1.v,m investigations 28 can be p1anned based on fi el a Fecennai ssance eat a from nli!lil!iill~ previous 29 investigations and reviews of available documents. Possible sources of 30 information may include universities, consulting firms, and government 31 agencies. A detailed list of possible sources of information is given in 32 Regulatory Guide 1.132. D-6

NUMARCComments March 19, 1993 Une ln/Une Out 1 D.2.3 Detailed l.!:!!t'I!:!:~ Investigations Te GhaneteFi2:e SeisfRie SeYFees 2 3 The following methods are suggested but they are not all-inclusive 4 and investigations should not be limited to them. Some procedures will 5 not be applicable to every site, and situations will occur that require 6 investigations that are not included in the following discussion. It is 7 anticipated that new technologies will be available in the future that will be 8 applicable to these investigations. -~ 11 12 D.2.3.1 Surface Investigations Surface exploration needed to assess i.!6., Reetectonic ee,uHtieFts ef P@Q:;,:m,::::::,e the geology of the aFea aF8YFtd the -s;li te lrii i S dependent on the 13 site location and may be carried out with the use of any appropriate 14 combination of geological, geophysical, seismological, and geotechnical 15 eng i nee r i ng tech ni que s IYllI::::1,1:::::::ti:1:1111:1::ir::11::1IiilM::11::::::es11::,11:111:1:xiiJ::;:1::11 . 16 17 D.2.3.1.1. Geological interpretations of aerial photographs and other 18 remote-sensing imagery, as appropriate for the particular site conditions, to 19 assist in identifying rock outcrops, faults and other tectonic features, frac- -~ 22 23 ture traces, geologic contacts, lineaments, soil conditions, and evidence of landslides or soil liquefaction. D.2.3.1.2. Mapping of topographic, geologic, geomorphic, and hydrologic 24 features at scales and contour intervals suitable for analysis, stratigraphy 25 (particularly Quaternary), surface tectonic structures such as fault zones, 26 and Quaternary geomorphic features. For offshore sites, coastal sites, or 27 sites located near lakes or rivers, this includes topography, geomorphology 28 (particularly mapping marine and fluvial terraces), bathymetry, geophysics 29 (such as seismic reflection), and hydrographic surveys to the extent needed 30 for evaluation. 31 32 0.2.3.1.3. Identification and evaluation of vertical crustal movements D-7

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 by (1) geodetic land surveying to identify and measure short-term crustal 2 movements (Refs. ~ID and il,D) and (2) geological analyses such _as analysis of 3 regional dissection and degradation patterns, marine and lacustrine terraces 4 and shorelines, fluvial adjustments such as changes in stream longitudinal 5 profiles or terraces, and other long-term changes such as elevation changes 6 across lava flows (Ref. 4§0). 7 8 0.2.3.1.4. Analysis of offset, displaced, or anomalous landforms such -~ 11 12 as displaced stream channels or changes in stream profiles or the upstream migration of knickpoints (Refs. eiD--10:tl\D), abrupt changes in fluvial deposits or terraces, changes in paleochannels across a fault (Ref. 91,i!D), or uplifted, downdropped, or laterally displaced marine terraces {Ref. -l-Gf!l!D). 13 14 D.2.3.1.5 . . Analysis of Quaternary sedimentary deposits within or near 15 tectonic zones such as fault zones and including: (1) fault-related or fault-16 controlled deposits, including sag ponds, graben fill deposits, and colluvial 17 wedges formed by the erosion of a fault paleoscarp and (2) non-fault-related, 18 but offset deposits including alluvial fans, debris cones, fluvial terrace, 19 and lake shoreline deposits. D.2.3.1.6. Identification and analysis of deformation features caused 22 by vibratory ground motions, including seismically induced liquefaction 23 features {sand boils, explosion craters, lateral spreads, settlement, soil 24 flows), mud volcanoes, landslides, rockfalls, deformed lake deposits or soil 25 horizons, shear zones, cracks, or fissures {Refs. Hl,llD and ~flD>. 26 27 0.2.3.1.7. Estimation of the ages of fault displacements by analysis of 28 the morphology of topographic fault scarps associated with or produced by sur-29 face rupture. Fault scarp morphology is useful in estimating age of last dis-30 placement, approximate size of the earthquake, recurrence intervals, slip 31 rate, and the nature of the causative fault at depth {Refs. ~Jl,D-~Jl!D). 32 D-8

NUMAAC c.omments March 19, 1993 Une ln/Une Out 1 D.2.3.2 Seismological Investigations 2 3 D.2.3.2.1. Listing all historically reported earthquakes lli:Hm 4 Mid!l::1,:11.:::::11r111:1: j!l!!!Uil!ioi.:~::,1::rt1.Ml'lt!Jjffiiti!ff['!tlh!iiifll!li§U.i.J.f'!ii!:tllt!iitffii§IJ!llli 1 1 5 iililiil:IIHii:Ii.FltiqtfilIIt.i.!ti:il that can reasonably be associated with seismic 6 sources, any part of which is within a radius of~ ggg km (~ 1!11 miles) of 7 the site :(i1~'-i:!!!l~i~:'-'-f:~IH9P]1, including date of occurrence and the following 8 measured or estimated data: highest intensity, magnitude, epicenter, depth, 9 focal mechanism, stress drop, etc. Historical seismicity includes both eo historically reported and instrumentally recorded data. For pre-instrumentally recorded data, intensity should be converted to magnitude, the 11 12 procedure used to convert it to magnitude should be clearly documented, aAd 13 e~ieeAteFs sheuld be deteF~iAed based eA iAteAsity eeAteuFs. Methods to 14 convert intensity values to magnitudes in the central and eastern U.S. are 15 described in References HlJ:D?Ill8D--14iOD. ..........,............ ******* 16 17 D.2.3.2.2. Seismic monitoring _in the site area that is established as 18 soon as poss i bl e after s i te sel ect ion . l1n::1:::1!i:l11I::::i:1:::::::111::::::111:[::::::1::::I1J:is.J::1:11:i11 19 linimli@:!lriniiltiiriilffilinl::::::11i:1riPi:!l!:1i::1::~:::rl!I:!:ilm11:111i:ii!ilimti::111:11J::111t]!!i 20 !liIIliiliii![llll@!tl:::::1:t:nil:~il!l:i:eiJ!@!l[1lliiil:!:!Iili!ii:!:IiisHlliillilliiiPlitillieil!l1!1! e 1 22 lie!!iiil!!I!!!l!ll!l)Ill! Il!Il!:(: t:1 ::::ml:lt:!1i!iii~fl!i!ii!I!il@t:i:1:11@:: 111 1 1 1 23 IIII:!irlmi:i.l[!pµ!;pp!iil:i::!illil!:ili!~l:1:,,:,::::::m11::1::1irilii!llli!l!II:Ili!l 24 25 1:r::iirnt:i§nt:1:i:11:1::1!::,::::I1:11t:::::111:i:::::I11::::::1ili!}!!i[!lili!lllii1:1i1t::::111mtl:iil!liil 1 26 1.:1rlfi84.!llli::;:1:11::t1ii!l:Uili:I:ilnli)ffif\1!!!l!I!r111!n:1!:t:t1.111:1s 27 inii!i lli[i!l!i§riniiI!::1111:::::::111,1::::::1,11:nil!i!liii:J.!rol!iiilil!iilisiilllsl 1 28 iiJi:16.§Uikii.!Ili!il.1 H1!I!t.nif!:i.!liti:l¥:,:;.ifnittv:;.: 1 29 30 t[:!I::!I!I![IPE:t1:1::11:jtd:1ji:jn1t:1i1t1.:ri6t:M:tl::~I~I:il!t::l:!!!IIII:i#:i:tirm1::iim 1 1 1 31 ii!1i!iml!el ii]:lffilfmiif!]J.!ill!slt!iIIi!iPill!!i!!I!liil!llllii!!I!ilitii1!in]!lill 1 32 lil:11[i!l!i:~:1mi1::~:* 1 1 D-9

NUMARCComments March 19, 1993 Une ln/Une Out 1 Moid=i,d.bbjiiihd.uld{,bj:ilin,1::fl,atiiffaip.i:jo.JitH:,aJ:ye1ars.:fjrFiotffJbiie6rii$.mitliljnfJj)\}a 2 ti11.l:@1m:j;pgjJ:;j;:j)j@l:m:::)~l)ilJ:11Rt:::i1ti~11:i::,;[£,t!l:f:fil!El,1:fa:,jljjygjm:11111!::!~1l:l:e1]n1.:,j:j,jJ,nl:l,1:!.!)f:,gn 3 sl!Iiliiil:iiiiiii:i:siJ! 4 5 0.2.3.3 Subsurface Investigations 6 Subsurface investigations in the 5'1,ite ,alrea &P !iii~: within the l'egieA 7 M:!:li:::i:'l:!:S~::11!!1¥ to identify and define seislRegeAic ii:i::11!:I sources and capable 8 tecteAic seln-ces l'li:]:;;,1 may include: -~ 11 12 0.2.3.3.1. Geophysical investigations such as air or ground magnetic and gravity surveys, seismic reflection and seismic refraction surveys, borehole geophysics, and ground-penetrating radar. 13 14 0.2.3.3.2. Core borings to map subsurface geology and obtain samples 15 for testing such as age dating. 16 17 D.2.3.3.3. Excavating and logging trenches across geological features 18 as part of the neotectonic investigation and to obtain samples for age-dating 19 those features. -~ 22 23 At some sites, deep soil, bodies of water, or other material may obscure geologic evidence of past activity along a tectonic structure. In such cases, the analysis of evidence elsewhere along the structure can be used to evaluate 24 its characteristics in the vicinity of the site (Refs. ~J.:IJD and ~l!IP>* 25 26 D2.4 Age-Dating 27 28 An important part of the geologic investigations to identify and define 29 potential seismic sources is the age-dating of geologic materials. The 30 following techniques are useful in dating Quaternary deposits. D-10

NUMAACC,omments March 19, 1993 Une ln/Une Out 1 D2.4.1 Radiometric Dating Methods 2 3

• Carbon 14 for dating organic materials (upper limit ranges from 4 30,000 up to 100,000 years) (Ref. ~11D).

5

• Potassium argon for dating volcanic rocks ranging in age from 6 about 50,000 to 10 million years (Ref. ~~D).

7

• Uranium series using the relative properties of various decay 8 products of 238U or 235U. Ages range from 10,000 to 350,000 (Ref.

-~ 11 12

         ~IID). 235U/ 238U can yield between 40,000 and 1,000,000 years (Ref. ~l~D).

Fission track using minerals such as zircon and apatite, with fissionable uranium in volcanic rocks. Although some 13 interpretation is required in counting tracks, the technique has 14 no inherent age range limitations if suitable materials are 15 available (Ref. uggo). 16

• Thermoluminescence (TZ) is best used for stratigraphic correlation 17 and determining relative ages rather than absolute ages. The 18 maximum age is 10 million years (Ref. lillD).

19

• Electron spin resonance (ESR) is used to date quartz that formed

-~ 22 23 in fault gouge during the fault event (Ref. ~il.D). D2.4.2 Other Quantitative Numerical Methods 24

• Paleomagnetic dating requires material containing magnetic-25 suscept~ble minerals with sufficient stratigraphic and time ranges 26 to provide several reversals. An independent time datum for 27 correlation with the polarity time scale is required (Ref. l-:12D).

28

• Thicknesses of weathering rind development on the margins of 29 clasts, such as caused by obsidian hydration, can be used to 30 estimate the age of deposits (Ref. -2425D).

31

• Cation-ratio dating of desert varnish on rock surfaces by chemical 32 analysis (Ref. SiiD).

D-11

NUMAACO>mments March 19, 1993 Une ln/Une Out 1

• Tephrochronology, which is the identification and correlation of 2 undated and dated volcanic ashes by geochemical and petrographic 3 analyses (Refs. Ul!D and ~1,1,D).

4

• Amino-acid racemization, which uses organic material and is based 5 on time-dependent diagenetic conversion of one form of amino-acid 6 polymer structure to another (Refs. ~IID and ~IQD).

7

• Lichenometry is used to estimate ages from sizes of lichens 8 growing on gravel or boulders (such as glacial deposits) (Ref.

9 ~1:10>

• ea 11
• Soil profile development is used to determine age based on measured amounts of accumulated pedogenic materials (Ref. 6-1-IZD).

12

• Dendrochronology is used to determine the ages of trees that were 13 affected by a tectonic event or other phenomena such as 14 landsliding or flooding (Refs. ~IID-64,i,))D).

15 16 D2.4.3 Relative Age-Dating Methods 17 18

• The relative degree of soil profile development of Band C 19 horizons can provide at least an order of magnitude estimate of 20 the ages of buried soils or relict surface soils on surficial e1 deposits (Refs. Hl!ID and ~l!§D). For B horizons, the diagnostic 22 characteristics include thickness, depth, amount, texture, type of 23 clay, soil structure and color, and amount of Fe oxides or Fe-Al-24 organic accumulation (Ref. 1-:lfD). For C horizons, the important 25 diagnostic characteristics are thickness, depth, stage of 26 development, and amount of pedogenic carbonate and other soluble 27 salts (Refs. 3e;j;J)D and a+IID). Other references for this subject 28 include References 6819.D through 42:13.!D.

29 30

• The relative degree of weathering of surface and subsurface clasts 31 in sedimentary deposits such as glacial moraines is useful but 32 requires independent means of age calibration (Ref. a:IID).

D-12

NUMAACComments March 19, 1993 Line In/Line Out I In the SCR it may not be possible to ffliiiri,ib1ly. demonstrate, ;,, aA 2 ahselyte maRAer, the age of last activity of a tectonic structure. In such 3 cases the NRC staff will accept association of such structures with geologic 4 structural features or tectonic processes that are geologically old (at least 5 pre-Quaternary) as an age indicator in the absence of conflicting evidence. 6 7 These investigative procedures should also be applied, where possible, 8 to characterize offshore structures (faults or fault zones, as well as folds, 9 uplift, or subsidence related to faulting at depth) for coastal sites or those ea 11 sites located adjacent to landlocked bodies of water. Investigations of off-shore structures will rely heavily on seismicity, geophysics, and bathymetry 12 rather than conventional geologic mapping methods that can !:!!IB!iJ:ill be used 13 effectively onshore. However, it is often useful to investigate similar 14 features onshore to learn more about the significant offshore features. 15 16 D2.5 Distinction Between Tectonic and Nontectonic Deformation 17 18 Nontectonic deformation, like tectonic deformation i!IIl:]i'i:11, can pose a 19 substantial hazard to nuclear power plants, but there are likely to be dif-20 ferences in the approaches used to resolve the issues raised by the two types a22 of phenomena. Therefore, non-tectonic deformation should be distinguished from tectonic deformation at a site. In past nuclear power plant licensing 23 activities, surface displacements caused by phenomena other than tectonic 24 phenomena have been confused with tectonically induced faulting . Such 25 features include faults on which the last displacement was induced by 26 glaciation or deglaciation; collapse structures, such as found in karst 27 terrain; and growth faulting, such as occurs in the Gulf Coastal Plain or in 28 other deep soil regions subject to extensive subsurface fluid withdrawal . 29 30 Glacially induced faylts l11lJl!~::1.9. generally doii not represent a deep-31 seated seismic or fault displacement hazard because the conditions that 32 created them are no longer present . However, residual stresses from D-13

NUMAFIC Comments March 19, 1993 Une ln/Une Out 1 Pleistocene glaciation may still be present in glaciated regions although they 2 are of less concern than active tectonically induced stresses. These features 3 should be investigated with respect to their relationship to current in situ 4 stresses. 5 6 The nature of faults related to collapse features can usually ~e defined 7 through geotechnical investigations and either can be avoided or, if feasible, 8 adequate engineering fixes can be provided. 9 ea 11 Large naturally occurring growth faults such as found in the coastal plain of Texas and Louisiana can pose a surface displacement hazard, even 12 though offset most likely occurs at a much less rapid rate than that of tec-13 tonic faults. They are not regarded as having the capacity to generate damag-14 ing earthquakes, can often be identified and avoided in siting, and their 15 displacements can be monitored. Some growth faults and antithetic faults 16 related to growth .faults are not easily identified; therefore, investigations 17 described above with respect to capable tectenic faults and fault zones should 18 be applied in regions where growth faults are known to be present. Local 19 human-induced growth faultj[ijgs. can be monitored and controlled or avoided. 20 91 If questionable features cannot be demonstrated to be of non-tectonic 22 origin they should be treated as tectonic deformation. 0-14

N~Conments March 19, 1993 Une ln/Une Out 1 REFERENCES 2 3 ifD.

,..-.:.; ,1:19.;J.pt¢Nl@l@llill£iilitirnltlil!)l1;MUidlll;.w11:l:11JtgtJJu1rd.

4 Hit.tii11J:i§i;If:ji]'1N.fiiitliijiiliiinlil.ii:fiiii:ii:Y.n:1:1B;j[j:§ji~ii1fXi!ltli::;::m1:it 5 Ri¥Ml:l~ii:Iil§.l::1lU:IIHliU:=1=1111tBlil~?.l+/-li;: 6 7 International Atomic Energy Agency, "Earthquakes and Associated 8 Topics in Relation to Nuclear Power Plant Siting," Safety Series No . 50-SG-Sl, Revision 1, 1991. -0 9 11 R. Reilinger, M. Bevis, and G. Jurkowski, "Tilt from Releveling: 12 An Overview of the U.S. Data Base," Tectonophvsics, Vol. 107, p. 13 315-330, 1984. 14 15 R. K. Mark et al., "An Assessment of the Accuracy of the Geodetic 16 Measurements that Led to the Recognition of the Southern 17 California Uplift," Journal of Geophvsical Research, Volume 86, 18 pp. 2783-2808, 1981 . 19 20 T. K. Rockwell et al., "Chronology and Rates of Faulting of tl1 Ventura River Terraces, California," Geological Society of America 22 Bulletin, Volume 95, pp. 1466-1474, 1984. 23 24 K. E. Sieh, "lateral Offsets and Revised Oates of Prehistoric 25 Earthquakes at Pallett Creet, Southern California," Journal of 26 Geoohvsical Research, Volume 89, No. 89, pp. 7641-7670, 1984. 27 28 K. E. Sieh and R.H. Jahns, "Holocene Activity of the San Andreas 29 Fault at Wallace Creek, California," Geological Societv of America 30 Bulletin, Volume 95, pp . 883-896, 19e4. 31 32 k. E. Sieh, M. Stuiver, and 0. Brillinger, "A More Precise D-15

NUMARC c.ommems March 19, 1993 Une ln/Une Out 1 Chronology of Earthquakes Produced by the San Andreas Fault in 2 Southern California," Journal of Geophysical Resea~ch, Volume 94, 3 pp . 603-623, 1989. 4 5 R. J. Weldon, III, and K. E. Sieh, "Holocene Rate of Slip and 6 Tentative Recurrence Interval for Large Earthquakes on the San 7 Andreas Fault, Cajon Pass, Southern California," Geological .~8 11 Society of America Bulletin, Volume 96, pp. 793-812, 1985. F. H. Swan, III, D. P. Schwartz, and L. S. Cluff, "Recurrence of Moderate to Large Magnitude Earthquakes Produced by Surface 12 Faulting on the Wasatch Fault Zone," Bulletin of the Sei~mological 13 Society of America, Volume 70, pp. 1431-1462, 1980. 14 15 Pacific Gas and Electric Company, "Final Report of the Diablo 16 Canyon Long Term Seismic Program; Diablo Canyon Power Plant," 17 Docket Nos. 50-275 and 50-323, 1988. 1 18 19 S. F. Obermeier et al ., "Geologic Evidence for Recurrent Moderate -~ 22 23 2;1:ao to Large Earthquakes Near Charleston, South Carolina," Science, Volume 227, pp. 408-411, 1985. D. Amick et al., "Paleoliquefaction Features Along the Atlantic 24 Seaboard," U.S. Nuclear Regulatory Commission, NUREG/CR-5613, 25 October 1990. 26 27 28 R. E. Wallace, "Profiles and Ages of Young Fault Scarps, North-29 Central Nevada," Geological Society of America Bulletin, Volume 30 88 , pp. 1267-1281, 1977. 1 31 Available for inspection or copying for a fee at the NRC Public Document 32 Room, 2120 L Street NW., Washington, DC. D-16

NUMARCc.omments March 19, 1993 Une In/Un* Out 1 R. E. Wallace, "Discussion--Nomographs for Estimating Components 2 of Fault Displacement from Measured Height of Fault Scarp," 3 Bulletin of the Association of Engineering Geologists, Volume 17, 4 pp. 39-45, 1980. 5 6 R. E. Wallace, "Active Faults, Paleoseismology, and Earthquake 7 Hazards: Earthquake Prediction--An International Review," Maurice 8 Ewing Series 4, American Geophysical Union, pp. 209-216, 1981. -~ 11 12 A. J. Crone and S. T. Harding, "Relationship of Late Quaternary Fault Scarps to Subjacent Faults, Eastern Great Basin, Utah," Geology, Volume 12, pp. 292-295, 1984. 13 14 0. W. Nuttli, "The Relation of Sustained Maximum Ground 15 Acceleration and Velocity to Earthquake Intensity and Magnitude, 16 State-of-the-Art for Assessing Earthquake Hazards in the Eastern 17 United States," U.S. Army Corps of Engineers Misc. Paper 5-73-1, 18 Report 16, 1979. 19 -~ 22 23

   -1&1:ID.

R. L. Street and F. T. Turcotte, "A Study of Northeastern North America Spectral Moments, Magnitudes and Intensities," Bulletin of the Seismological Society of ' America, Volume 67, pp. 599-614, 1977. 24 25 R. L. Street and A. Lacroix, "An Empirical Study of New England 26 Seis- micity," Bulletin of the Seismological Society of America, 27 Volume 69, pp. 159-176, 1979. 28 29 H. Rood et al ., "Safety Evaluation Report Related to the Operation 30 of Diablo Canyon Nuclear Power Plant, Units 1 and 2," USNRC, 31 NUREG-0675, Supplement No . 34, June 1991. 32 D-17

NUMARC C,ommen1S March 19, 1993 Une ln/Une Out 1 J. F. Callender, "Tectonics -and Seismicity," Chapter 4 in 2 "Techniques for Determining Probabilities of Event~ and Processes 3 Affecting the Performance of Geologic Repositories," NUREG/CR-3964 4 (SAND 86-0196), Volume 1, Edited by R. L. Hunter and C. J. Mann, 5 pp. 89-125, June 1989. 6 7 D. R. Muhs and B. J. Szabo, "Uranium-Series Age of the Eel Point 8 Terrace, San Clemente Island, California," Geology, Volume 10, pp. .0-9 11 23-26, 1982. M. Ikeya, T. Miki, and K. Tanaka, "Dating of a Fault by Electron 12 Spin Resonance on Intrafault Materials," Science, Volume 215, pp. 13 1392-1393, 1982. 14 15 S. M. Colman and K. L. Pierce, "Weathering Rinds on Andesitic and 16 Basaltic Stones as a Quaternary Age Indicator, Western United 17 States," Professional Paper 1210, U.S. Geological Survey. 1981. 18 19 R. I. Dorn, "Cation-Ratio Dating: A New Rock Varnish Age- -~ 22 23 Determination Technique," Quaternary Research, Volume 20, pp. 49-73, 1983. P. D. Sheets and D. K. Grayson, eds., Volcanic Activity and Human 24 Ecology, Academic Press, New York, 1979. 25 26 D-18

NUMARCC,omments March 19, 1993 Line ln/Une Out 1 mao.

      ...... S. Self and R. J. s. Sparks, eds . , "Tephra Studies," Proceedjngs 2            of the NATO Advanced Studies Institute, Tephra Stu~ies as a Tool 3            jn Quaternary Research, D. Reidel Publ. Co., Dordrecht, Holland, 4            1981.

5 6 J. L. Bada and P. M. Helfman, "Amino Acid Racemization Dating of 7 Fossil Bones," World Archeology, 1975. -~ 8

   ~Ibo.
     ....... J. L. Bada and R. Protsch, "Racemization Reaction of Aspartic Acid and its Use in Dating Fossil Bones," Proceedings of the National 11            Academy of Science, Volume 70, pp. 1331-1334, 1973 .

12 13 W. W. Locke, J . T. Andrews, and P. J. Webber, "A Manual for 14 Lichenometry," T~chnical Bulletin 26, British Geomorphological 15 Research Group, University of East Anglia, Norwich, 1979. 16 17 3+3.20.

     , ...... M. N. Hachette, "Dating Quaternary Faults in the Southwestern 18            United States by Using Buried Calcic Paleosols," U.S. Geological 19            Survey Journal of Research, Volume 6, pp. 369-381, 1978.

-~ 22 23 R. Page, "Dating Episodes of Faulting from Tree Rings: Effects of the 1958 Rupture of the Fairweather Fault on the Tree Growth," Geological Society of America Bulletin, Volume 81, pp. 3085-3094, 24 1970 . 25 26 K. E. Si eh, "Prehistoric Earthquakes Produced by Slip on the San 27 Andreas Fault at Pallett Creek, California," Journal of 28 Geophysical Research, Volume 83, pp. 3907-3939, 1978 . 29 30 0-19

NUMARCC,omments March 19, 1993 Une ln/Une Out 1 B. F. Atwater and D. K. Yamaguchi, "Sudden, Probably Coseismic 2 Submergences of Holocene Trees and Grass in Coastal Washington 3 State," Geology, Volume 19, pp. 706-709, 1991. 4 5 M. N. Hachette, "Soil Dating Techniques, Western Region (United 6 States)," Open-file Report OFR-82-840, U.S. Geological Survey, p. 7 137-140, 1982. 8 -09 11 12 L. D. McFadden and J. C. Tinsley, "Soil Profile Development in Xeric Climates: A Summary," in J.C. Tinsley, J.C. Matti, and L. D. McFadden, eds., Guidebook, Field Trip No. 12, Geological Society of America, Cordillera Section, pp. 15-19, 1982. 13 14 J. W. Hardin, "A Quantitative Index of Soil Development from Field 15 Descriptions: Examples from a Chronosequence in Central 16 California," Geoderma, Volume 28, pp. 2-18, 1982. 17 18 J. C. Matti et al., "Holocene Faulting History as Recorded by 19 Alluvial Stratigraphy Within the Cucamonga Fault Zone; A. -~ 22 23 Preliminary View," in J. C. Tinsley, J. C. Matti, and L. D. McFadden, eds., Guidebook, Field Trip No. 12, Geological Society of America, Cordillera Section, pp. 29-44, 1982. 24 P.A. Pearthree and S.S. Calvo, "Late Quaternary Faulting West of 25 the Santa Rita Mountains South of Tucson, Arizona," Masters 26 Thesis, University of Arizona, Tucson, AZ, 1982. 27 28 P. A. Pearthree, C. M. Menges, and L. Mayer, "Distribution, 29 Recurrence, and Possible Tectonic Implications of Late Quaternary 30 Faulting in Arizona," Open-file Report 83-20, Arizona Bureau of 31 Geology and Mineral Technology, Tucson, 1983 . 32 D-20

NUMARC C,omments March 19, 1993 Une ln/Une Out 1 E. A. Keller et al., "Tectonic Geomorphology of the San Andreas 2 Fault Zone in the Southern Indio Hills, Coachella ~alley, 3 California," Geological Society of America Bulletin,

• volume 93, 4 pp. 45-56, 1982.

5 6 0. A. Chadwick, S. Hecker, and J. Fonseca, "A Soils Chronosequence 7 at Terrace Creek: Studies of Late Quaternary Tectonism in Dixie Valley, Nevada," Open-file Report 84-0090, U.S. Geological Survey, -9 8 1984. D-21

NUMARC Q>mmenta March 19, 1993 Line In/Line Out 1 REGULATORY ANALYSIS 2 3 A separate regulatory analysis was not prepared for this 4 regulatory guide. The draft regulatory analysis, "Proposed Revision of 10 CFR 5 Part 100 and 10 CFR Part 50," provides the regulatory basis for this guide and 6 examines the costs and benefits of the rule as implemented by the guide. A 7 copy of the draft regulatory analysis is available for inspection and copying 8 for a fee at the NRC Public Document Room, 2120 L Street NW. (Lower Level), -9 Washington, DC, as Enclosure 2 to Secy 92-215. D-22

NUMAAC Ciomments March 19, 1993 Line In/Line Out 1 APPENDIX E 2 3 PROCEDURE FOR EVALUATION OF SEISMIC SOURCES IN THE 4 STABLE CONTINENTAL REGION 5 6 7 E.l Introduction 8 -1 9 Appendix B to 10 CFR Part 100 requires that site geological, 10 seismological, and geophysical investigations be conducted for nuclear power plant sites. For sites located in the Stable Continental Region (SCR) 12 (eastern U.S.), Appendix B permits the use of an accepted seismic hazard 13 methodology and data to determine the Safe Shutdown Earthquake Ground Motion 14 (SSE). On a site-specific basis, Appendix B requires that a demonstration be 15 made to show that the site investigation does not provide information that is 16 not adequately represented in the existing seismic sources and parameters. 17 This appendix describes an evaluation to provide this demonstration. 18 19 The objective of the evaluation is to integrate the results of the site 20 geological, seismological, and geotechnical investigations and the seismic 21 sources and parameters that have been accepted on a regional scale. 92 Completion of the evaluation described in this section and site-specific 23 acceptance of the existing seismic sources provides full integration of the 24 information produced by the site investigation. 25 26 The Staff has identified probabilistic seismic hazard (PSHA) methods and 27 data (seismic sources and parameters) that have been developed for the SCR 28 which are acceptable to perform a site-specific hazard analysis. Experience 29 in applying PSHA methods and hazard results in the SCR has demonstrated the 30 stability of the median annual probability of exceeding ground motions and 31 thus of the process to determine the SSE. The objective of the integrated 32 evaluation described in this appendix is to confirm this position by providing 33 a reasonable assurance that existing seismic source interpretations are not 34 inconsistent with the up-to-date information and its interpretation. Of E-1

NlJMAFC Ciommenta March 19, 1993 Line ln/LJne Out I particular importance is the necessity to reasonably assure that new 2 information, which is derived from the site investigation, does not constitute 3 a systematic change in expert interpretations and thus a systematic increase 4 in the median hazard curve.* Alternatively, the evaluation may conclude that, 5 on a site-specific basis, modifications of the existing seismic sources may be 6 required. This appendix provides guidelines to conduct the integrated 7 evaluation. 8 -~ 11 12 E.2 Background The Staff has accepted the use of seismic source characterizations developed for the SCR to estimate the seismic hazard to determine the SSE 13 (Ref. IE, 2E). An important part of the accepted PSHA methodologies was the 14 elicitation and quantification of the scientific interpretation of earth 15 science data regarding the future location and likelihood of occurrence of 16 earthquakes. The quantification of uncertainty includes the uncertainty 17 attributed to individual experts as well as the variability between experts. 18 Thus, there is a wide range of alternative seismic source characterizations 19 and seismicity parameters that are represented in these studies. -~ 22 23 When the EPRI seismic hazard methodology and data were submitted to the USNRC for review, it was requested that approval for use of the methodology and seismic source interpretations be given for a period of 5-10 years. It is 24 anticipated that at appropriate intervals, the existing seismic source 25 interpretations will be updated to insure that they continue to be an adequate 26 basis to perform seismic hazard evaluations in the SCR. 27 28 E.3 Approach 29 30 The approach to evaluate existing seismic sources for sites in the SCR 31 is based on the following: 32 E-2

NIJ.MRC Convnenta Match 19, 1993 Line lnfUne Out 1 Geological, geophysical, and seismological 2 investigations must be performed on a site-specific 3 basis as required by Appendix B to 10 CFR Part 100 and 4 described in Appendix D to this Regulatory Guide. 5 6 2. An accepted PSHA methodology and seismic source 7 .interpretations are an adequate basis to evaluate the 8 ground motion hazard, subject to a demonstration (confirmation) that information produced in the site ~ investigation is adequately represented. In selected 11 circumstances, modifica~ion of existing seismic 12 sources may be necessary. 13 14 The purpose of this evaluation is to integrate the results of the site 15 investigation with existing seismic source interpretations. 16 17 In the integrated evaluation, assessments are performed to determine 18 whether the results of the site investigation and possible seismic source or 19 seismicity parameter interpretations that may be derived from this information -~ are consistent with the data and interpretations made by the earth science experts when the accepted PSHA method and seismic sources were developed. The 22 evaluation is performed to determine whether new data or scientific 23 interpretations constitute a basis to conclude that the existing seismic 24 sources are not adequate to assess the SSE. 25 26 The scope of the integrated evaluation is divided into three levels. The 27 evaluation is performed in successive steps as a function of the necessity to 28 assess new information. In the first level, the consistency of the earth 29 science database that was available at the time the accepted seismic sources 30 were developed and the results of the site investigation is evaluated. In the 31 second level, the existing seismic source interpretations are examined to 32 assess whether the range of interpretations incorporates the new data and/or E-3

NUMAAC Col I uI NII ,ta March 19, 1993 Line In/Line Out 1 interpretations derived from tbe site investigation. In the third level, the 2 sensitivity of the site hazard to the new data or interpretations developed 3 from the site investigation is examined. 4 5 As part of the integrated evaluation an assessment must be made whether 6 new data are significant in terms of their impact on the determination of the 7 site hazard and the SSE. The approach used to assess the significance of new 8 information is described in Section E.6. & The evaluation described in this appendix assumes that an accepted PSHA 11 method is one that can be applied throughout the SCR and is adequately 12 documented. Documentation should describe the earth science database used 1n 13 the study and the basis for the expert seismic source interpretations and 14 pa~ameter assessments. Lacking full documentation, only Level 3 of the 15 evaluation can be performed. Levels 1 and 2 require that adequate 16 documentation be available to assess the consistency of the data and seismic 17 source interpretations with new, site-specific information. 18 19 The seismic hazard assessment performed as part of the seismic siting -~ 22 23 process is specifically conducted for the purpose of estimating the SSE ground motion. The evaluation process described in this appendix has been developed for the purpose of judging the stability of seismic hazard results at gr~und motion levels similar to the SSE of operating nuclear power plants (e.g., PGAs 24 in the range of 0.10 to 0.309.). Other applications of probabilistic seismic 25 hazard assessments, such as input to seismic probabilistic risk assessments 26 (seismic PRAs), require an assessment of the probability of exceedance of 27 ground motions much greater than the design basis (3 to 4 times the SSE). The 28 conclusions of the integrated evaluation do not necessarily apply to the 29 assessment of the ground motions significantly greater than the SSE. E-4

N ~ Commenta March 19, 1gga Une lnfUne Out 1 E.4 Site Geological. Se1smoJoq1caJ. and Geophysical Investigations 2 3 The identification and characterization of seismic sources*for a site 4 begins with a site-specific geological, seismological, and geophysical 5 *investigation. These investigations are outlined in Appendix D to this 6 Regulatory Guide in terms of three scopes of study: 7 8 1. Site Area - Within 8 kilometers of the site: Comprehensive, -~11 12 state-of-the-art data collection and analysis of the potential for permanent ground displacement and for the presence of seismic sources are conducted. The scope of these studies includes review of existing data, augmented by more detailed investigations conducted in the 13 specified area. 14 15 2.. Within 25 kilometers of the site: Reconnaissance-level investigations,

16 guided in part by the existing data base, augmented by detailed 17 investigations as needed to comprehensively evaluate potential seismic 18 sources are conducted within this specified distance from the site.

3. Within 200 kilometers of the site: Literature review augmented by reconnaissance-level investigations as appropriate to update and verify 22 the existing data b~se within this distance from the site are conducted.

23 24 The existing geological, seismological, and geophysical databases that were* 25 compiled and evaluated in the EPRI (Ref. IE) study constitute a current, 26 accepted, comprehensive, and detailed (at a scale of 1 or 2 kilometers) body 27 ~f data and interpretations on which to base these further site-specific 28 investigations. These further investigations should be performed in such a 29 manner as to provide a high-confidence database for the site. E-5

NUMAFC Comments March 19, 1993 Une ln/LJne Out I E.5 Identification and Characterization of Seismic Sources 2 3 Using the combination of the existing EPRI (Ref. IE) database-and the 4 additional data and interpretations derived from the site-specific 5 investigations, it is useful to consider what new data may be collected during

  .6 the detailed site investigations, and what previously unavailable information 7 or interpretations might be incorporated in an evaluation of accepted seismic 8 source interpretations. The following list and descriptions indicates seven 9 possibilities .

11 Mjcroearthauake Data. Continued recording of small earthquakes (M<3.0) 12 in_the SCR may provide regional data not previously available. Also, 13 according to Item 10 of Appendix D, seismic monitoring is expected to be 14 initiated in an area around the site. These data may suggest or 15 indicate the presence of a localized earthquake potential that was more

, 16       generalized previously. The microearthquake data may suggest or 17       indicate such specific parameters as depth range of seismicity, or
-18        extent (length or volume) of a localized source.

19 -1 20 22 23

2. M>5 Earthguake. One or more moderate to large earthquakes (M>5) may occur within the SCR in the next several years or decades. Since historic events of this size may not have been explicitly considered in the existing seismic source characterizations, they were likely 24 considered implicitly. They constitute important new data.

25 26 3. Geologic Evidence of Quaternary Activity. Further geologic 27 investigations within the SCR will, in certain localities, identify 28 geologic evidence of Quaternary faulting or strong earthquake motions. 29 Such evidence may include paleoliquefaction or surface faulting that 30 deforms or involves Quaternary or younger geologic strata. Data of this 31 type may provide evidence for the rate of occurrence of moderate to 32 large magnitude events, the extent of a seismic source, or the maximum E-6

NUMAFC Comments March 1s, 1m Line In/Line Out 1 magnitude earthquake. 2 3 4. Geophysical Evidence of Quaternary Seismic Activity. Aeromagnetic, 4 gravity, and seismic reflection studies within the SCR will continue to 5 identify crustal geophysical structures that are related to subsurface 6 and surface geological structures, and may be related to patterns and 7 mechanisms of microearthquakes and moderate earthquakes. While the 8 geophysical data may not provide direct evidence of Quaternary tectonic -~ 11 12 5. activity, the geologic structural interpretations and associations may provide indirect evidence of potential seismic source characteristics. New Concepts of Tectonics and Earthquake Source Definition. New ideas 13 and experts to espouse them will arise. Such ideas may incorporate data 14 not previously addressed, or they may reinterpret data already 15 considered in the accepted PSHA method. 16 17 6. Geodetic Measurements. The rapid expansion of satellite-based geodetic 18 networks will provide data and interpretations of rates and styles of 19 deformation in the SCR. In time, these networks will increase in -~ 22 23 station density and accuracy. It is reasonable to expect that some of the results will have implications for seismic source characterization in terms of boundaries and earthquake recurrence. 24 7. In-situ Stress Measurements. Subsurface data continue to be collected 25 and interpreted to assess levels and patterns of local and regional 26 tectonic stress. As such research continues, implications will be 27 developed on seismic source characterizations. 28 29 In future nuclear power plant siting investigations, the treatment of new data 30 and interpretations such as those described above should consist of one or 31 more evaluations of the significance of the new data in the context of 32 existing seismic source interpretations. This approach is preferable to the E-7

NUMARC Commenta March 19, 1gga Line In/Line Out I approach of developing a complete new set or sets of seismic source 2 interpretations. 3 4 Because of the comprehensive nature of the accepted PSHA methods and 5 seismic sources, it is necessary to only carry out an assessment to determine 6 whether there is new data or interpretations that are not adequately 7 represented in the existing seismic sources. The identification and 8 characterization of existing seismic sources is inherently stable and has been accepted by the Staff (Ref. 3E). Differences between data obtained from site & investigations and existing data need to be addressed only if the differences 11 are potentially significant. New data is considered significant if it 12 requires further evaluation, requires additional alternative seismic sources 13 and seismicity parameters, and results in an increase in the site seismic 14 hazard. 15 16 E.6 Evaluation of New Data and Interpretations 17 18 To evaluate the existing seismic source interpretations and the results 19 of the site-specific investigation, a three-level assessment is carried out. -~ 22 23 The levels of the evaluation are performed in order to assess:

• consistency of the earth science database used in the EPRI study and the results of the site investigation, 24 25
• the sensitivity of the existing seismic source 26 interpretations and alternative interpretations based 27 on site-specific information, and 28 29
• the sensitivity of the median seismic hazard curve 30 based on existing seismic sources to alternative 31 seismic source or seismicity parameter 32 interpretations.

E-8

NUMAAC C.ornmenta March 1g, 1;93 Un* In/Line Out I Figure E-1 provides a flow diagram of the three-levels of the integrated 2 evaluation. 3 4 The seismic hazard information base developed as part of the hazard 5 asses~ment can be used to evaluate new data or alternative seismic source 6 interpretations (see Appendix C to this Regulatory -Guide). The infonnation 7 base provides quantitative data on the contribution of seismic sources and 8 magnitude-distance pairs to the SSE and ~hus can be used to provide insight to -~ 11 12 the relative importance of new data. E.6.1 Significance of New Data 13 The integrated evaluation is performed in successive levels as further 14 assessment is required to consider the consistency of the existing seismic 15 sources with the site information. At the completion of each level, if it is 16* concluded that the existing seismic sources are not inconsistent, or are 17 conservative with respect to the information of the same type, developed in 18 the site investigation, then it is not necessary to proceed to the next level 19 (see Fig. E-1). For example, a Level 1 evaluation may conclude that the site -~ 22 23 investigation has not produced new information that contributes to the characterization of seismic sources. In this case, the basis for the EPRI* seismic sources is not challenged and therefore no further evaluation (i.e., development of alternative seismic sources) is required. 24 25 An important step in the integrated evaluation is to assess whether new 26 information is significant in terms of the determination of the SSE. 27 Assessment of whether new data is significant can be made ~teach level of the 28 evaluation. As general guidance, new data are not considered significant if: 29 30

• after examination they require no further evaluation, 31 32
• they require no alternative seismic sources or seismicity 33 parameters, or 34 35
• they result in maintaining or decreasing the site median hazard.

E-9

MJMAACCommenta March 1g, 1~ Line In/line Out 1 The significance of new information can be evaluated in terms of the impact on 2 the determination of the SSE. The third criterion listed above states if the 3 site median hazard is maintained or decreases, new data is not considered 4 significant. The extensive seismic hazard evaluations that have been 5 performed in the SCR quantify the uncertainty in the probability of exceedance 6 of ground motion.- Considering the uncertainty in the seismic hazard 7 evaluations, it is reasonable to anticipate that information from a site 8 investigation could lead to small increases or decreases in the median hazard. Such increases are, from an engineering perspective, not unacceptable if the & resulting change in the SSE is small. Based on seismic design practice and the margins available in nuclear power plant designs (Ref. 4E) variation of 11 12 the SSE of not more than twenty percent is acceptable. This variation is 13 within the inherent variability in the estimation of structure and equipment 14 response to earthquake ground motions and much less than the seismic margins 15 in nuclear power plant design. 16 17 Given a limit on the variation in the SSE, there is a corresponding 18 limit to a change in the median hazard curve. The tolerance for increase in 19 the median hazard depends on the slope of the hazard curve between the SSE and -~ 22 23 the value that is twenty percent higher. The limiting change, denoted as a Hazard Factor (HF) is defined by, HF -= a1. 2 24 25 where a is the slope, on a log-log scale, of the median hazard curve, as 26 derived from the existing seismic sources. The HF defines the limiting 27 .multiplicative increase in the median hazard curve. The HF can be used in the 28 Level 1, 2, and 3 evaluations to assess the relative importance of new data. E-10

NUMARC Commenta March 19, 1993 Une lnfUne Out 1 E.6.2 Evaluation Levels 2 3 Level I; Evaluation of New Data in Terms of the Existing Data Base 4 5 As part of the EPRI seismic hazard study, a comprehensive earth science 6 database was compiled and provided to the Earth Science Teams (Ref. IE). In 7 addition, the EPRI teams documented the information and basis for their 8 seismic source interpretations. A comparable database and expert -~ 11 12 documentation was not compiled in the LLNL study. Consequently, the Level 1 evaluation can only be conducted for the EPRI study. The first evaluation to be performed is an assessment of whether new 13 data collected for a site are consistent with the comparable data in the 14 existing EPRI database. In this comparison, differences are assessed in such 15 - areas as spatial patterns of seismicity, deformation rates, relationships to 16 significant past earthquake activity, etc. Quantitative significance tests 17 may be performed as appropriate to objectively identify the presence of 18 differences. It is typical that new data sets will be found to be consistent 19 with the previous comparable data sets used to establish the EPRI seismic -~ 22 23 source characterizations. If so, then these existing characterizations can be used for the site evaluation. If the new data sets are not consistent with the data sets used in the 24 EPRI analysis and they require further evaluation, then the Level 2 assessment 25 should be performed. 26 27 28 Level 2: Evaluation of New Data and Interpretations in Terms of the Range of 29 EPRI Interpretations 30 31 For new data or new interpretations or concepts that require further 32 evaluation, it is necessary to assess if they are consistent with the range of E-11

M.IMAFCCommenta March 19, 1993 Une ln/Une Out 1 existing seismic source interpretations. This determination is made by 2 assessing the implications of the new data or interpretation for such source 3 characterization elements as alternative source geometry, alternative maximum 4 magnitudes, or alternative recurrence rates or models. In most cases, the new 5 data interpretations, will already be incorporated within the range of 6 alternatives, and thus will not affect the current seismic source 7 characterizations. 8 Level 3: Evaluation of New Data and Interpretations in Terms of the SSE 11 A Level 3 evaluation is performed if further evaluation of new data or 12 interpretations is required. In this assessment the sensitivity of the SSE to 13 alternative seismic source characterizations based on new data or 14 interpretations is examined. 15 16 If there is no difference in the median hazard or the SSE (as described 17 in Section E.6.1), then the new data or interpretations are not considered 18 significant. If however, after the Level 3 evaluation, the new data or 19 interpretations are found to be significant, then a formal, site-specific update of the seismic sources that is consistent with the original accepted PSHA methodology (Refs. IE, 3E) must be made. 22 23 E.7 Procedure 24 25 This section outlines the procedure for conducting the evaluation of the 26 accepted PSHA methodology for sites in the SCR. 27 28 Level 1: - Evaluation of New Data 1n Terms of EPRI Data Base 29 30 In this evaluation, which is performed for the EPRI database only, a 31 systematic review is conducted of the earth science data obtained from the 32 site investigation. The review focusses on two basic issues. First, has new E-12

NUMARC C,ommenta Match 1g, 1993 Un* In/Line Out 1 data been obtained that is different from information that was available in 2 the EPRI database? When new data does exist, the question now focuses on 3 whether the data is significant (requires further evaluation) with respect to 4 the identification and characterization of seismic sources. An affirmative 5 conclusion requires that a Level 2 evaluation be performed. 6 7 Examples of significant new data include: 8 -~ 11 12

1. Data that provides evidence of a previously unknown tectonic feature capable of generating earthquakes of magnitude 5 or greater.

13 2. - New evidence that indicates that the spatial 14 dimensions of known tectonic features are larger than 15 was previously considered and requires further 16 evaluation. 17 18 3. Data that suggests that the rate of earthquake 19 occurrences of 1t1i,>5.0 within 100 km of the site is -~ 22 23 4. higher and requires further evaluation. New data or methods of estimation that impact the assessment of the maximum magnitude earthquake and 24 requires further evaluation. 25 26 5. Availability of new data or methods to model the 27 occurrence of earthquakes in the seismic hazard 28 analysis .. For example, models to account for fault 29 rupture in the hazard analysis. 30 31 32 E-13

NJMAFICCommenta March H~, 1993 Une ln/Une Out 1 As part of this evaluation, a concise tabular su1T111ary should be prepared that 2 identifies the following: 3 4

• EPRI earth science data category 5

6

• Information available in the EPRI data base 7

8

• Results of the site-specific investigation ao9 11
• Assessment of the information in this category relative to the need for further evaluation 12 13 This summary will identify any new information that requires further 14 evaluation in terms of the identification and characterization of seismic 15 sources in the vicinity of a site.

16 17 Level 2; Evaluation of New Data and Interpretations in Terms of the Range of 18 Seismic Source Interpretations 19 -~ 22 23 This evaluation is conducted if the EPRI and site-specific data require further evaluation to assess their impact on the identification and characterization of seismic sources. 24 New data are considered significant in the Level 2 evaluation if the 25 following conditions are satisfied: 26 27 28 a} New data are not accounted for in the existing 29 seismic sour~e or seismicity parameter interpretations 30 and require further interpretations, and E-14

M.A1AAC Commenta March 19, 1993 Une ln/LJne Out 1 b) the estimated median earthquake occurrence rate 2 {II\Z5,0 events) within 100 km of the site is 3 sig~ificantly different {exceed) than the rate 4 estimated by the existing seismic source 5 interpretations. 6 7 2. Known or discovered tectonic features require finite- .~8 11 fault rupture be considered in the estimation of ground motion. In the Level 2 evaluation, an assessment is made to determine whether 12 the existing seismic source interpretations are consistent with the results of 13 the site investigation. The steps in this evaluation should: 14 15 Identify the seismic sources used in the hazard analysis. 16 Surrunarize the expert characterization of each seismic source, 17 including the tectonic feature(s) being modeled, the probability 18 of being active, maximum magnitude, etc. 19 -~ 22 23

2. Assess the seismic source zone characteristics (source geometry and probability of activity, maximum magnitude, and earthquake occurrence rates) in the context of significant new data or ne*w interpretations.

24 25 As part of this evaluation, it may be necessary to evaluate the sensitivity of 26 earthquake occurrence rates within the vicinity of a site to variations in 27 seismic source characteristics. 28 29 A comparison between the,existing seismic source characterizations and 30 alternative characterizations suggested by new data should focus on 31 differences in the occurrence of earthquakes within 100 km of the site. If 32 the difference between the median rate of occurrence of m125.0 events derived E-15

tfJMARC Comments March 19, 1993 Un* In/Un* Out 1 from the existing seismic sources and the alterative parameters is.less than 2 the HF, it can be concluded that the new data and interpretations are not 3 significantly different. If this is not the case, a Level 3 evaluation must 4 be perfonned. 5 6 Where new data suggest that the maximum magnitude may be greater, a 7 Level 3 evaluation will generally be required because the effect of larger 8 earthquakes on ground motions depends on the distance to the site. ~ Level 3: Evaluation of New Data and Interpretations in Terms of the SSE 11 12 In this step the sensitivity of the SSE to seismic source 13 interpretations or seismicity parameters based on new data is assessed. The 14 sensitivity of the median hazard to alternative seismic source 15 characterizations, is evaluated in the following steps: 16 17 Step 1. Perform the seismic hazard assessment for spectral 18 acceleration for 10, 5, 2.5, and 1 Hz based on the 19 alternative seismic sources or seismicity parameters. -~ 22

- 23 Step 2.      Detennine the median probability of exceedance for the average spectral accelerations, 5-10 Hz and 1-2.5 Hz.

24 Step 3. From the median hazard curves for the average spectral 25 acceleration from Step 2, determine the ground motion 26 corresponding to the Reference Probability (see Appendix B 27 of this Regulatory Guide). 28 29 Step 4. If the ground motions determined in Step 3 do not exceed the 30 SSE based on the existing seismic sources by more than 31 twenty percent, no further evaluation is required. 32 E-16

NUMAFC Corrmenta March 19, 1993

                                                                          ~ In/Line Out I Ground Motion Attenuation Models 2

3 The seismic ha.zard ca lcul at i ans that are performed should use the ground 4 motion attenuation models that were used in the seismic hazard study at the 5 time {Ref. IE, 3E). If alternative ground motion models are used, it must be 6 demonstrated that the median seismic hazard results for average spectral 7 acceleration at 5 and 10 Hz and I and 2.5 Hz are consistent with the results .~8 that would be obtained using the models from the original study. Alternatively a reevaluation of the seismic hazard at all operating nuclear power plant sites as of [Effective Date of the Final Rule] could be conducted 11 to determine the appropriate Reference Probability {see Appendix B). E-17

NJMAfCCommenta March 19, 1993 Line ln/Une Out 1 , REFERENCES 2 . . 3 IE. Electric Power Research Institute, *seismic Hazard Methodology for the 4 Central and Eastern United States,* Volumes 1 through 10, EPRI NP-4726-5 A, 1988. 6 7 2E. D. L. Bernreuter et al., "Seismic Hazard Characterization of. 69 Nuclear 8 Plant Sites East of the Rocky Mountains," NUREG/CR-5250*, January 1989. -~ 11 12 3E. U.S. Nuclear Regulatory Commission, "Safety Evaluation Review of the SOG/EPRI Topical Report Titled Seismic Hazard Methodology for the Central and Eastern United Sates,R EPRI NP-4726-A, Sept. 20, 1998. 13 14 4E. Electric Power Research Institute, "A Methodology for Assessment of 15 Nuclear Power Plant Seismic Margin,R Revision 1, EPRI NP-6041, August 16 1991. E-18

NUMARC C:Omments March 19, 1993 Line In/Line Out 1 2 3 4 5 6 7 8 9 10 Site and Site Region Geological, Geophysical EPRI 11 and Seismological Earth Science Databuc 12 Investigation 13 14 ~7 18 19 20 21 22 23 Conduct Seismic 24 Huard Analysis 25 Seismic Hazard 26 Information Base 27 28 29 30 S~pecific Update of Seismic Sources 31 C34 35 36

                                      '                DetmnineSSE 37 38 39 40       Figure E-1 Flow diagram of the integrated evaluation for existing 41       seismic source interpretations for the SCR.

E-19

NUMARC C,omments March 18, 1993 Une ln/Une Out 1 instrumentation data and for determining whether plant shutdown would be 2 required by the proposed amendments to 10 CFR Part 50. 3 Any information collection activities mentioned in this draft regulatory 4 guide are contained as requirements in the proposed amendments to 10 CFR Part 5 50 that would provide the regulatory basis for this guide. The proposed 6 amendments have been submitted to the Office of Management and Budget for . 7 clearance that may be appropriate under the Paperwork Reduction Act. Such 8 clearance, if obtained, would also apply to any information collection -~ 11 12 activities mentioned in this guide. B. DISCUSSION 13 When an earthquake occurs, ground motion data are recorded by the seismic 14 instrumentation. 1 . These data are used to make an early determination of the 15 degree of severity of the seismic event. The data from the seismic 16 instrumentation, coupled with information obtained from a plant walkdown, are 17 used to make the initial determination of whether the plant should be shut 18 down, if it has not already been shut down by operational perturbations 19 resulting from the seismic event. If on the basis of these initial evalua-tions (instrumentation data and walkdown) it is concluded that the plant shut-down criteria have not been exceeded, it is presumed that the plant will not 22 be shut down. Guidance is being developed on postshutdown inspections and 23 plant restart; see Draft Regulatory Guide DG-1018, "Restart of a Nuclear Power 24 Plant Shut Down by a Seismic Event. 11 25 The Electric Power Research Institute has developed guidelines that will 26 enable licensees to quickly identify and assess earthquake effects on nuclear 27 power plants. These guidelines are in EPRI NP-5930, "A Criterion for Deter-28 mining Exceedance of the Operating Basis Earthquake," July 1988; EPRI NP-6695, 29 "Guidelines for Nuclear Plant Response to an Earthquake," December 19892 ; and 30 EPRI TR-100082, "Standardization of Cumulative Absolute Velocity," December 31 1991. 2 2 32 EPRI reports may be obtained from the Electric Power Research Institute, 33 Research Reports Center, P.O. Box 50490, Palo Alto, CA 94303 2

NUMARCComments March 18, 1993 Une ln/Une Out 1 This regulatory guide is based on the assumption that the nuclear power 2 plant has operable seismic instrumentation, including the equi~ment and soft-3 ware required to process the data within 4 hours after an earthquake. This is 4 necessary because the decision to shut down the plant will be made, in part, 5 by comparing the recorded data against OBE exceedance criteria. The decision 6 to shut down the plant is also based on the results of the pl=ant walkdown 7 inspections that take place within 8 hours of the event. If the seismic 8 instrumentation is inoperable, the guidelines in Appendix A to this guide .09 11 would be used to determine whether the operating basis earthquake ground motion (OBE) has been exceeded. Shutdown of the nuclear power plant would be required if the vibratory 12 ground motion experienced exceeds that of the OBE. Two criteria for determin-13 ing exceedance of the OBE are provided in EPRI NP-5930: a threshold response 14 spectrum ordinate criterion and a cumulative absolute velocity criterion 15 (CAY). A procedure to standardize the calculation of the CAY is provided in 16 EPRI TR-100082. A spectral velocity threshold has also been recommended by 17 EPRI since some structures have fundamental frequencies below the range speci-18 fied in EPRI NP-5930. The NRC staff now recommends 1.0 to 2.0 Hz for the 19 range of the spectral velocity limit since some structures have fundamental frequencies below 1.5 Hz . The former range was 1.5 to 2.0 Hz. The NRC staff does not endorse the philosophy discussed in EPRI NP-6695, 22 Section 4.3. 4 (first paragraph, last sentence), pertaining to plant shutdown 23 considerations following an earthquake based on the need for continued power 24 generation in the region . If the licensee determines that plant shutdown is 25 required by the Commission's regulations, but the licensee does not consider 26 it prudent to do so, the licensee may ask for an emergency exemption from the 27 requirements of the regulation pursuant to 10 CFR Part 50.12 so that the plant 28 need not shut down if the exemption is granted. 29 Appendix B to this guide provides definitions to be used with this 30 guidance . 31 32 C. REGULATORY POSITION 33 34 1. BASE-LINE DATA 3

NUMARC Comments March 18, 1993 Une ln/Une Out I I.I Information Related to Seismic Instrumentation 2 3 A file containing information on all the seismic instrumentation should 4 be kept at the plant. The .file should include: 5 6 I. Information on each instrument type such as make, model, and serial 7 number; manufacturers' data sheet; list of special features or options; per-8 formance characteristics; examples of typical instrumentation readings and -~ 11 12 interpretations; operations and maintenance manuals; repair procedures (manu-facturers' recommendations for repairing common problems); and a list of any special requirements, e.g., maintenance, operational, installation. 13 2. Plan views and vertical sections showing the location of each 14 seismic instrument and the orientation of the instrument axis with respect to 15 a plant reference axis. 16 17 3. A complete service history of each seismic instrument. The service 18 history should include information such as dates of servicing, description of 19 completed work, and calibration records and data (where applicable). 22 cumulative absolute velocity (see Regulatory Position 4). These data should 23 be obtained after the initial installation and each serviciflg of the free-24 field instrumentation using a suitable earthquake time-h i story (e .g. , the 25 October 1987 Whittier, California, earthquake) or manufacture's calibration 26 standard. 27 28 1. 2 Planning for Postearthquake Inspections 29 30 The selection of equipment and structures for inspections and the content 31 of the baseline inspections as described in Sections 5.3. 1 and 5.3 . 2. 1 of EPRI 32 NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake," are accept-33 able to the NRC staff for satisfying the proposed requirements in Paragraph 4

NUMAAC Comments March 18, 1993 Une ln/Une Out 1 IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 for ensuring the safety of 2 nuclear power plants. 3 4 2. IMMEDIATE POSTEARTHOUAKE ACTIONS 5 6 The guidelines for invnediate postearthquake actions specified in Sections 7 4.3.1 and 4.3 .2 (including Section 5.3.2.1 and items 7 and 8 of Table 5-1) of 8 EPRI NP-6695 are acceptable to the NRC staff for satisfying the requirements 9 proposed in Paragraph IV(a)(3) of Proposed Appendix S to 10 CFR Part 50. ea11 3. EVALUATION OF GROUND MOTION RECORDS 12 13 3.1 Data Identification 14 15 A record collection log should be maintained at the plant, and all data 16 should be identifiable and traceable with respect to: 11 18 1. The date and time of collection, 19 20 2. The make, model, serial number, location, and orientation of the - instrument (sensor) from which the . record was collected. 22 23 3.2 Data Collection 24 25 3.2.1 Only personnel trained in the operation of the instrument should 26 collect the data. 27 28 3.2.2 The steps for removing and storing records from each seismic 29 instrument should be planned and performed in accordance with established 30 procedures. 31 32 3.2.3 Extreme caution should be exercised to prevent accidental damage 33 to the recording media and instruments during data collection and subsequent 34 handling. 5

NUMARC c.ommenta March 18, 1993 Une ln/Une Out 1 .3...:.1..:.! As data are collected and the instrumentation is inspected, notes 2 should be made regarding the condition of the instrument and its installation, 3 for example, instrument flooded, mounting surface tilted, fallen objects that 4 struck the instrument or the instrument mounting surface. 5 6 .3..:_Ll For validation of the collected data, a reference signal (see 7 Regulatory Position 1.1(4)) should be added to the record without affecting 8 the previously recorded data. -~ 11 12 3.2.6 If the instrument's operation appears to have been normal, the instrument should remain in service without readjustment or change that would defeat attempts to obtain postevent calibration. 13 14 3.3 Record Evaluation 15 16 Records should be analyzed according to the manufacturer's specifications 17 and the results of the analysis should be evaluated . Any record anomalies, 18 invalid data, and nonpertinent signals should be noted, along with any known 19 causes. -~ 22 23

4. DETERMINING OBE .EXCEEDANCE The evaluation to determine whether the OBE was exceeded should be 24 performed using data obtained from the three components of the free-field 25 greund motion (i.e., two horizontal and one vertical). The evaluation may be 26 performed on uncorrected earthquake records. It was found in a study of 27 uncorrected versus corrected earthquake records (see EPRI NP-5930) that the 28 use of uncorrected records is conservative. The evaluation should consist of 29 a check of the response spectrum, cumulative absolute velocity limit, and the 30 oper ab i l i ty of the i nst rumen tat ion . 11:~::1:::::::111i:1:,11:11t:J:iliil!lltli:i.liI:i:11:1.11:::I11::11:j:n 31 liin:iffiiiiii!iliigliiiiiiifii:iiii:iiniliiili;:iiii!iijiis~faj 32 33 4.1 Response Spectrum Check 34 6

March 18, 1993 Line ti/Line Out l The QBE response spectrum is exceeded if any one of the three components 2 (two horizontal and one vertical) of the 5 percent damped free-field gPOYAd 3 mot 1on response spectra 111.:111:11R:IIJ.il.h1'!It.a& miii.B.i:~:11:i1: gi,iix:Jii.ii.li:Iiiiil 4 ~p~,-~::: :...:f~::=:J.::e:1~:r,j:~~~i:P.r::~9-~l*t:NJ.f.@!-ni:?~t::,n.~?::~P::?~!~, i s 1arge r than :

    =

5 6 1. The corresponding design response spect~al acceleration (QBE 7 spectrum if used, otherwise 1/3 of the safe shutdown earthquake 8 (SSE) spectrum) or 0.2g, whichever is greater, for frequencies 9 between 2 to 10 Hz, or eo 11 2. The corresponding design response spectral velocity (QBE spectrum if 12 used, otherwise 1/3 of the SSE spectrum) or a spectral velocity of 6 13 inches per second, whichever is greater, for frequencies between 1 14 to 2 Hz . 15 16 4.2 Cumulative Absolute Velocity {CAY} Limit 17 18 For each component of the free-field gPeYAd motion, the CAY should be 19 calculated as follows: (1) the absolute acceleration (g units) time-history 20 is divided into I-second intervals, (2) each I-second interval that has at e1 least I exceedance of 0.025g is integrated over time, (3) all the integrated 22 values are summed together to arrive at the CAY. The CAY limit is exceeded if 23 any CAY calculation is greater than 0.16 g-second. Additional information on 24 how to determine the CAY is provided in EPRI TR-100082. 25 26 4.3 Instrument Operability Check 27 28 After an earthquake at the plant site, the response spectrum and CAY 29 should be obtained using the calibration standard (see Regulatory Position 30 1. 1( 4)) to demonstrate that the S;YStem 1:~;~:::1:11,:,,,,i::1:!11niJl!iillii:::::1,11ur1:l:!:!i.14. 31 nafflt#:n.l was functioning properly. 32 33 4.4 Inoperable Instrumentation 34 7

NUMMCC!lmmMII March 18, 1993 Une ln/Une Out 1 If the seismic instrumentation is inoperable, the criteria in Appendix A 2 to this guide should be used to determine whether the OBE has been exceeded. 3 4 5. CRITERIA FOR PLANT SHUTDOWN 5 6 If the OBE is exceeded or significant plant damage occurs, the plant must 7 be shut doWn-:-jUtfffll)ii.ill'lnl::#1ii1illj}:ga1.:1:i.rilffilillifflillR]lli)Ili.Q.l,:UJ,n)IS.t:ll[:-,1:, a 11:1:1t!¥:fJ::;::1tiiiiiili1 9 eo 5.1 OBE Exceedance 11 12 If the response spectrum check and the CAV limit (performed in accordance 13 with Regulatory Positi on! 4.1 and 4. 2) were exceeded, Ill! the OBE was 14 exceeded and plant shutdown is required . If either limit does not exceed the 15 criterion, the earthquake motion did not exceed the OBE. Jl1lJgp]J[¥j)j)J:jlilJB.IIP!l.t!!i 1 15 11,,1r@m::::::~:1:m~:11:::::1,:::i:11@::::::111::::::111:1:,:1i:1,n::::::,it:sli.s111:;.:::::::1:1,iI1111r::::a:,:::;:1,:,m.1::::::11:i11 1 11 !lill!llifflilIIli1jilil1!!1IIilillilt1::::::1nJ:!tli1ifPillbi:iill1Rillllif1§i!ll@:t The 18 determination of whether or not the QBE has been exceeded should be performed 19 even if the plant automatically trips off-line as a result of the earthquake . 20 5.2 Damage e1 22 The plant should be shut down if the walkdown inspections, performed in 23 accordance with Regulatory Position 2 (SeetieA 4.3.2 ef EPRI NP &&9§), 24 di scover damage 1,:1111,:11:1:::::]l[i:::1:11:1:::::::11ml§IIM!iiliil!l!iiI!l!i:1IJ:[l-25 26 6. PRE-SHUTDOWN INSPECTIONS 27 28 The pre-shutdown inspecti ons described in Section 4.3.4 of EPRI NP-6695, 29 "Guidelines for Nuclear Plant Response to an Earthquake," 1 with the last sen-30 tence in the first paragraph of Section 4.3 . 4 deleted, are acceptable to the 31 NRC staff for sat i sfying the requ i rements proposed i n Paragraph IV{a){3) of 32 Proposed Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear power 33 plants . 8

NUMARCCommenta March 18, 1993 Line In/Line Out 1 The following paragraph in Section 4.3.4 of EPRI NP-6695 is repeated to 2 emphasize that the plant should shut down in an orderly manner. 3 4 "Prior to initiating plant shutdown following an earthquake, visual 5 inspections and control board checks of safe shutdown systems should 6 be performed by plant operations personnel, and the availability of 7 off-site and emergency power sources should be determined. The pur-8 pose of these inspections is to determine the effect of the earth~ -~ 11 12 quake on essential safe shutdown equipment which is not normally in use during power operation so that any resets or repairs required as a result of the earthquake can be performed, or alternate equipment can be readied, prior to initiating shutdown activities. In order 13 to ascertain possible fuel and reactor internal damage, the follow-14 ing checks should be made, if possible, before plant shutdown is 15 initiated . . . . " 16 17 If the OBE was not exceeded and. the walkdown inspection i!nl!lili! 18 illll:1Bli:i]if:linfilfi@!:::i:11n1111ili1::ii11iirrinii.[ilil:liisil:lil:i:i:Iti1[§IRIIIIB1§ill 19 indicates no damage to the nuclear power plant, shutdown of the plant is not required. The plant may continue to operate (or restart following a post-trip review, if it tripped off-line because of the earthquake}. 9

NUMARC Comments March 18, 1993 Une ln/Une Out 1 D. IMPLEMENTATION 2 3 The purpose of this section is to provide guidance to applicants and 4 licensees regarding the NRC staff's plans for using this regulatory guide. 5 This draft guide has been released to encourage public participation in 6 its development. Except in those cases in which the applicant proposes an 7 acceptable alternative method for complying with the specified portions of the 8 Conrnission's regulations, the method to be described in the active guide 9 reflecting public conments will be used in the evaluation of applications for eo 11 construction permits, operating licenses, combined licenses, or design certi-fication submitted after the implementation date to be specified in the active 12 guide. This guide would not be used in the evaluation of an application for 13 an operating license submitted after the implementation date to be specified 14 in the active guide if the construction permit was issued prior to that date.

NUMARC Q>mmenta March 18, 1993 Une In/Un* Out 1 APPENDIX A 2 INTERIM OPERATING BASIS EARTHQUAKE EXCEEDANCE GUIDELINES 3 4 Th i s regulatory gu ide is based on the assumpt ion that the nuclear power 5 plant has operable seismic instrumentati on . If the seismic instrumentation is 6 inoperable, the following should be used to determine whether the operating 7 basis earthquake ground motion (OBE) has been exceeded : 8 9 1. For plants at wh i ch instrumentally determined data are available &Aly at - 0 the fe YAdat i BA l eve l 1":1:::::JJ'.r@ffl!:::::i§JJiJ:::1i[il,~!\-!!:J:JJJ:J~j,j:i:i1J!:Jii~:]~n,ji!Ji:1.::::::111:lf,1,~IJq9 11 iti.l.liffl:1:sn:::;:si,iiit;lllltlli.ifi[illll@iil:t:12:::::r.u.m,::::::11,,1;:EJ1illllil1:nillilill, the 12 cumulative absolute veloc ity (CAV) limit (see Regulatory Posit ion 4. 2 of 13 this guide) is not applicabl e, aAd a:f@fMHitilili!ltij!iiiFifij determination of 14 OBE exceedance i s based on the response spectrum check described in 15 Regulatory Position 4. 1 of this regulatory guide. A comparison is made 16 between the foundation-level design response spectra and data obtained 17 from the foundation-level instruments . If the response spectrum check at 18 any foundat ion is exceeded, the OBE is exceeded and shutdewA is warraAted 19 lll];IJ:ln!:l:Imllli:;IIJ::IilMli!wn

• 20 411 2. For plants at wh ich no instrumental data are avail able, the OBE wi ll be 22 considered to have been exceeded and 111::::::11::1:11::::::mii!]J!§@::::~::ibi!!l!l!l~§i~ shutdewA 23 te ~e warraAted if one of the following applies:

24 25 1. The earthquake resulted 1:1tilffl!IfRl!IIIIIlllii!!IJI!l[:'.f.llEB!:11111i!e):: 4ft 26 Medified Merealli IAteAsity (MMI) VI BF greateF withiA &klR ef the 27 plaAt, 28 29 2. The earthquake was 1,m111i:11:1:,r:1t.Hil]iliio.~::::::1.tm:r,1~ of magn itude 6. 0 or 30 greater, or 31 32 3. The earthquake was felt within the plant and was of magn itude 5.0 or 33 greater and occurred with in 200 km of the pl ant . 34 A- 1

NUMARC Comments March 18, 1993 Une ln/Une Out 1 3. A postearthquake plant walkdown should be conducted (see Regulatory 2 Position 2 of this guide). 3 4 4. If plant shutdown is warranted under the above guidelines, the plant 5 should be shut down in an orderly manner (see Regulatory Position 6 of 6 this guide). 7 .~ 8 The determinations of epicentral locationT iija magnitude, aR~ iRteRsit~ by the U.S. Geological Survey, National Earthquake Information Center, 11 will usually take precedence over other estimates; however, regional and 12 local determinations will be used if they are considered to be more 13 accurate. Also, higher quality damage reports or a lack of damage 14 reports from the nuclear power plant site or its inrnediate vicinity will 15 take precedence ~ver more distant reports. 16 17 18 19 -~ 22 23 24 A-2

NUMARCQJmments March 18, 1993 Une ln/Une Out 1 APPENDIX B 2 DEFINITIONS 3 4 Design Response Spectra . Response spectra used to design Seismic Category I 5 structures, systems , and components . 6 7 Operating Basis Earthquake Ground Motion (OBE) . The vibratory ground motion 8 f'.oF whieh those f'.eatYFes of the RYeleaF po*1eF plant neeessal"Y foF eontinYed 9 opeFation withoYt YndYe Fisk to the health and safety of the pYhlie will eo Femain fynetional. The ¥a1Ye of the OB£ is set hy the applieant iiii~J.!ili~ 11 11:11::::::,1:1,1:IlifiillBilil1nl:!:1lilniilelll!in::::::1n1::11:,:::::::111,:1::11:11:1[l:1Ml:i@lli;ilill![l.¥iflii 1 12 iiil:lilliil:!Iiil:IliI:iliil1ilnI!Iliiil

• 1 13 14 Spectral Accelerat i on . The accelerat i on response of a l i near oscillator with 15 prescribed. frequency and damp i ng.

16 17 Spectral Velocity . The velocity response of a linear oscillator with pre-18 scri bed frequency and damping . 19 20 e1 22 23 24 25 26 27 28 29 30 B-1

NUMAAC Commems March 18, 1993 Une ln/Une Out 1 REGULATORY ANALYSIS 2 3 A separate regulatory analys;s was not prepared for th;s regulatory 4 guide. The draft regulatory analys;s, "Proposed Revisions of 10 CFR Part 100 5 and 10 CFR Part 50," prov;des the regulatory basis for this gu;de and examines 6 the costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analys;s is available for inspection and copy;ng for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, -~ 11 12 as Enclosure 2 to Secy 92-215. 13 14 15 16 17 18 19 -~ 22 23 24 25 26 28 29 30 RA-1

NUMARCC.ornmems March 18, 1993 Line In/Line Out Enclosure 9 October 1992 Division 1 Draft OG-1016

Contact:

R. M. Kenneally (301) 492-3893 DRAFT REGULATORY GUIDE OG-1016 (Second Proposed Revision 2 to Regulatory Guide 1.12) (Previou~ly issued as Draft MS-140-5) NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES 1 A. INTRODUCTION 2 3 In 10 CFR Part 20, "Standards for Protection Against Radiation," 4 licensees are required to make every reasonable effort to maintain radiation 5 exposures as low as is reasonably achievable. Paragraph (c) of§ 50.36, -~8 9

   "Technical Specifications," to 10 CFR Part 50, "Domestic Licensing of Pro-duction and Utilization Facilities," requires the technical specifications of a facility to include surveillance requirements to ensure that the neces-sary quality of systems and components is maintained, that facility opera-10 tion will be within safety limits, and that the limiting conditions of 11 operation will be met. Paragraph IV(a)(4) of Proposed Appendix S, "Earth-12 quake Engineering Criteria for Nuclear Power Plants," to 10 CFR Part 50 13 would require that suitable instrumentation be provided so that the seismic 14 response of nuclear power plant features important to safety can be evalu-15 ated promptly. Paragraph IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 16 would require shutdown of the nuclear power plant if vibratory ground motion 17 exceeding that of the operating basis earthquake ground motion (OBE) 18 occurs.*

19 20 *Guidance is being developed in Draft Regulatory Guide DG-1017, "Pre-21 Earthquake Planning and Immediate Nuclear Power Plant Operator Post-22 Earthquake Actions," on plant shutdown Criteria. 1

NUMAFC Comments March 18, 1993 Line In/Line Out 1 This guide is being developed to describe seismic instrumentation i!!filg 2 !19.l]llll]llfl)t@Pililf!il]Illl:l!]!ll acceptable to the NRC staff for s~t i sfyi ng the 3 requirements of Parts 20 and 50 and the Proposed Appendix S to Part 50. 4 Any information collection activities mentioned in this draft regulatory 5 guide are contained as requirements in the proposed amendments to 10 CFR 6 Part 50 that would provide the regulatory basis for this guide. The proposed 7 amendments have been submitted to the Office of Management and Budget for 8 clearance that may be appropriate under the Paperwork Reduction Act. Such -~ 11 12 clearance, if obtained, would also apply to any information collection activities mentioned in this guide. B. DISCUSSION 13 14 When an earthquake occurs, it is important to i1111Rediately i:il~E]~rimPI 15 li.£1:i;ggf;)g assess the effects of the earthquake at the nuclear power pl ant. 16 State ef the arts lolid-state digital time-history accelerographs installed 17 at appropriate locations will provide time-history data on the seismic 18 response of the free-field, containment structure, and other Category I 19 structures. The instrumentation should be located so that a comparison and -~ 22 23 evaluation of such response may be made with the design basis and so that occupat i ona1 radiation exposures iliiiillililiiiil}i!l::flli:jiiiIIiii!lliiii are maintained as low as reasonably achievable (ALARA). Free-field instrumentation data would be used to determine whether the 24 OBE greYAd metieA has been exceeded (see Draft Regulatory Guide DG-1017). 25 Foundation-level instrumentation would provide data on the actual -seismic 26 input to the containment and other buildings and i,p~l:]jQi[flituim,tiia W&tH-4 27 quantify differences between the vibratory ground motion at the free-field and 28 foundation level. Instrumentation is not located on equipment, piping, or 29 supports since experience has shown that data obtained at these locations are 30 obscured by vibratory motion associated with normal plant operation. 31 The guidance being developed in Draft Regulatory Guide DG-1017 is based 32 on the assumption that the nuclear power plant has operable seismic instrumen-33 tation, including the equipment and software required to process the data 34 within 4 hours after an earthquake. This is necessary heeayse the deeisieR te 2

NUMAAC Canments March 18, 1993 Line In/Line Out 1 shwt ElewA the pl aAt ~*i 11 he RlaEle' i A ,a,t t i.iflil.i.ldlniWJ:lfillil~W:::1b.1.l:[dH6.:t=~::1 2 !liil!tiP:l!lJJ::1t11::1:Jlilim~:,:11:1:11:::::11:1::11::::B:;r,ilii by comparing the recorded

                                                                                             .       data        '

3 against OBE exceedance criteria,1&(61. lhe decisieA te shwt dewA the t>laAt is 4 alse haseEI eA the results of the e,e,ateP iJ=jiitfiwa l kdown inspections that take 5 place within 8 hours of the event. 6 It may not be necessary for identical nuclear power units on a given site 7 to each be provided with seismic instrumentation if essentially the same 8 seismic response at each of the units is expected from a given earthquake. 11 12 An evaluation of seismic instrumentation operational experience noted that instruments have been out of service during plant shutdown and sometimes during plant operation. The instrumentation system should be operable at all times. If the seismic instrumentation l!iii!iir.Jtil§nf!lli::!:!1iliimlinilJ:ei:!tilt:111 13 is inoperable, the guidelines in Appendix BI to Draft Regulatory Guide DG-14 1017 would be used to determine whether the OBE has been exceeded. 15 Instrumentation characteristics, installation, activation, remote indica-16 tion, and maintenance are described in this guide to help ensure (1) that the 17 data provided are comparable with th~ data used in the design of the nuclear 18 power plant, (2) that exceedence of the OBE can be determined, and (3) that 19 the equipment will perform as required. The Appendix to this guide provides definitions to be used with this guidance. 22 23 C. REGULATORY POSITION 24 25 The type, locations, operability, characteristics, installation, 26 actuation, remote indication, and maintenance of seismic instrumentation 27 described below are acceptable to the NRC staff for satisfying the require-28 ments in 10 CFR Part 20, 10 CFR 50.36(c), and Paragraph IV(a)(4) of Proposed 29 Appendix S to 10 CFR Part 50 for ensuring the safety of nuclear power plants. 30 31 1. SEISMIC INSTRUMENTATION TYPE AND LOCATION 32 3

NUMAFCComments March 18, 1993 Une ln/Une Out 1 Ll State ef the aFt s Solid-state digital instrumentation that will 2 enable the processing of data at the plant site within 4 hours of the seismic 3 event should be used. 4 5 Ll A triaxial time-history accelerograph should be provided at each of 6 the following locations: 7 8

• 11 12 2.

3. Containment foundation.

                  -lw&   99:!     elevation, (excluding the foundation) on a structure 13                 internal to the containment.

14 15 4. -lw& Iii independent Category I structure foundation, (for 16 instance, the diesel generator building -aftEi Ir the auxiliary 17 building) where the response is different from that of the 18 containment structure. 19

5. An elevation (excluding the foundation) on each ef the

'22 23 24 6. independent Category I structure, selected in 4 above. If seismic iselateFs aFe usee, instFumentatien sheule ~e places en ~eth the Figie ane iselatee peFtiens ef the stFuctuFes at 25 appFeKimately the same ele¥atiens. 26 :111:111::::::1inlliiiitl:l:!iii!Ji.1t1:11:fil::1::f::1:111::,::::::1111isr.1::::::1:::::::1iil!ijilnl 27 ili!Iiliii!liili['ilsiiiiIIlb.it:1:~:1,:r111,:::::111111~::::IliilliY!ilifil:1111: 1 28 11:::::=11.1::H~i{,i,i:;.:t:tl1.ii!I!Hji.t]rwlo.iii.Hji.{Jl.it':l:ii.11.:1.::11:1::i1:1.10.1:~:1,1.a::::=:w, 1 29 :11111:::: :1::::::im::::::ai:11;:r111::i,11111:1r.1i: 1 30 31 1.3 The specific locations for instrumentation should be determined by 32 the nuclear plant designer to obtain the most pertinent information consistent 33 with maintaining occupational radiation exposures ALARA for the location, 34 installation, and maintenance of seismic instrumentation. In general: 4

NUMAFCc.omments March 18, 1993 Une In/Un* Out 1 1.3.1 A design review of the location, installation, and 2 maintenance of proposed instrumentation for maintaining exposures ALARA should 3 be performed by the facility in the planning stage in accordance with 4 Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational 5 Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably 6 Achievable." 7 8 1.3. 2 Instrumentation should be placed in a location with as low 9 a dose rate as is practical, consistent with other requirements. eo 11 1.3.3 Instruments should be selected to require minimal 12 maintenance and in-service inspection, as well as minimal time and numbers of 13 personnel to conduct installation and maintenance. 14 1s 1,:,~,:1,~=:1=:i:::::rn11:::::::~1:~~:1~:@l::1:::::~:1@n:11~1i::1HP1!:P:!:!:!l!::::::~~:n:,:j,@:l::1::11::::!::111::=::~=1::esij,11:::::::1:e 16 l:bl~::::::,,~:1::~:?~..,~rq~~::::;~9-~:J:pq:!:::1~:m1:::::!:p~::::::~~:::::1,~~::1:1r=~~~~i::::!:lnli§li:i:i:iP.~::1:::~~!::::::~~l~S~:~::!::::'-~i~ 17 lri[lllBl:ltB:Jij::11::::J:l!lrlii:i:::::::11:l:iii:::::::11111111:1:;::::::Illi:11:1:m,1::I::111::::::1,11111111::::::1j::J::1i::11 18 j.§J!tlli.@Imfiilllttl::li;iif:lmt11m1,1.!m:itll!l:i111ttln1!I:1.n11J.:1r:fimil:1.,11:::111::::::1;111111:1ttiII§Yi! 1 19 elll:@fl::~1m11iitl:111::::::11,:inl!:11:,1.mg:1:11:11111::~:::Ii:::::1snliill1J:lfltft:!:!itlll:ll11lnli:l!:1 20 ~*1n111r:1!t::::m~1:j::i;111::::::1111et1::::::mx:1~1M::~:~!1:1@1:r11:~,1m1:,:1i:1::1t~m::111:ni::::111,:~1!::::1111:11m e 1 llltffllfiYl!$:;@an;:r1;rt::::t.b.lfiJIIUld'111:1.tf::::1rmlb.l.f:11d.nl.lU~mt.il@!Jll.ml<<t,U:t1;.:ttt:l::au~i::11illill 22 s11:1=1:11n1t:i:Rtll)j:l:=:1:l!!!Pl::1::11:P:1l:~Jmt::P:ftm@i:1:1n1:wHt!l111:J.::1:1::11sqn::t1.t:r111:1::t:!4l!Mt:::1:rn 23 ;61:b$.tftJ.Pt$.:b1ritt'P.l:l¢.ld:n11.:rn,=h1n=1:11:c.1tt:~:nB.:J::#,t@il¢.~:t=~l.:1w:n1.fiei1*10.a@m$.Uttl1t.1¥IdfflA 24 §lfilllii:Yf'~:::ti:n:tJQgg=b1~n::1tn~:111,n:oettlll1.J!iJ.!l@iJiifli)$.l!tV.Mtttmm~,1~J.:i'.1:1v1m1:t1il: 25 pffiUtit:ttil1Itf:edtttfdttMiafthjttiliHffiot:ld,~im)ir.e%t.irii)fitifitffiilitlihttiemBiili§.'fbjffin 26 w.1:1:1t:ttiitue::::::,111.11,nc.<<:::::,r1::1,r1:d1:m::u111:1e1:t:n1et:11.11!;:t::r1t111:i~1,1:,,::::::1.or 21 Pe.t1:rmi:n.i:n1mtEx1,1t1n1.1.::::::rit1m:::i11nfflir1t~:n1mm1,,:1::$.n:1,tt11u1~-w~:hw.11.mie.tilPP'-*t 28 29 IWIV§=rl)M:11i:il:fi:J.:M+/-ltiiitttti@'.j~¢illltAl1Jlib.1.Ill:ll1.l9.l.J.:tU:ftllfflJAfiMtml 30 lllltMtl:llili:ll:n~1::::::11:11!tlllinll:H:t:1:1t:tti1t::1:111tu:11r:1111:11M:lP.tr!!!ltltlI:ltUtJ:n1:Mt 1 31 Wb~Je.t:1:na#.~t\!J.¢.t.U.Jtf.thlh$.l.~lt.l.$.{j$.l*q1:1::::::l.m:t1lJ.¢.t.d::::::1t:tllt'.t4,tiJijh$,J:=1b.'.1t.mhate.t:tl~n 32 ffi8:I.Jtl@l:i:1iI:m1l!Ili:l::t1::1~;:::;:,:1:::=:=111:i111:1:1:~bllU11mis:i11111mt:1:~:1:1!1:11,,:~:;::::111:::;:m11:1:lr@I. 33 m,1ii:1tr1::11111111!t11:111111::111,1::::::11::::::11::111111::::::111111,1:::::11:11i::111:::::1111:,1:::::::11ii111l::::::1i1 1 34 ~:i!!:::::ff:i::!m11.t:1s,1i:1:rggnt1Hrn:1ng9::J.:l.:tn@1wa:11:es111g@:=tatt:1t:!&1n11r.1::::::11ms1Yn1.J:: 1 5

NUMARCc.omments March 18, 1993 Une In/UM ().rt 1 11tMJ&b.ittbffilt;Httt1t;ft(itf:lB.iifllRl/$.t::1:m,1!$/@l.dfUtll{UH:ili*::m:b.O.'i::li:U11::::::d,nandi 2 !111,:J:: 3 4 2. INSTRUMENTATION AT MULTI-UNIT SITES 5 6 Instrumentation in addition to that installed for a single unit will not 7 be required if essenti ally the same seismic response is expected at the other 8 units based on the seismic analysis used in the seismic design of the plant. However, if there are separate control rooms, annunciation should be provided to both control rooms as specified in Regulatory Position 7. 11 12 3. SEISMIC INSTRUMENTATION OPERABILITY 13 14 The se i smi c i nstrumentat ion should operate during al l modes of plant 15 operation, including periods of plant shutdown . The maintenance and repair 16 procedures should provide for keeping the maximum number of i nstruments i n 17 service during plant operation and shutdown . 18 19 4. INSTRUMENTATION CHARACTERISTICS 20 e1 Ll The design shoul d include prov i sions for i n-service test i ng . The 22 i nstruments should be capable of periodic channel checks during normal plant 23 operation . 24 25 4.2 The instruments should have the capability for in-place functional 26 test i ng . 27 28 4.3 The instFumentatien en the feundatien and at cle¥atiens ,1ithin the 29 same building OF strueturc sheuld be inteFeenneeted fer eelllfRon starting and 30 E8RIIIOn timing, and the 1 Instrumentation li:l!I::1m1ilil:::::1:11illUi!)!:!mlni.1:iii!i11]§J.1 31 iiii,: shoul d contai n provisi ons for liliI!Iiisirl:~:ar:11:t:;,:::::::11siii:~:11:1:tJ!:ililJ:91 32 i.~,q~n,:::::::~:iij~~:ijµ~!P.~~~~::~ij:::::::J:'-q#.l)!l:i::]~~q~:~:11!::::;:~~~ an extern a1 remote a1arm to 33 i ndicate actuat i on . 34 6

NUMARC Comments March 18, 1993 Une ln/Une Out 1 4.4 The pre-event memory of the instrumentation should be sufficient to 2 record the onset of the earthquake; for example, it should have the ability to 3 record the 3 seconds prior to seismic-trigger actuation. It should operate 4 continuously during the period in which the earthquake exceeds the seismic-5 trigger threshold and for a minimum of 5. seconds beyond the last seismic-6 trigger signal. The instrumentation should be capable of a minimum of 25 7 minutes

• of continuous recording iHf'@JJlr:tii,if11:e.J!lf:l@lii.lid.ltm!tt:b.et:l'.t:nlllFt.l 8 m1i1v:liiII§lliliilltBl11:ei
• 11 12 4.5 Acceleration Sensors 4.5.1. The dynamic range should be 1000:1 zero to peak, for 13 example, 0.00lg to l.0g.

14 15 4.5.2. The frequency range should be 0.20 Hz to 50 Hz, or an 16 equivalent demonstrated to be adequate by computational techniques applied to 17 the resultant accelerogram. 18 19 4.6 Recorder 4.6.1. The sample rate should be at least 200 samples per second 22 i:,::::::11m1::::::11:::::1:11::::::1:i,,1:::::p:i:r11"i:~:Pni-23 24 4.6.2. The bandwidth should be at least from 0.20 Hz to 50 Hz. 25 26 4.6.3. The dynamic range should be 1000:1. 27 28 4.7 Seismic Trigger. The actuating level should be adjustable .f.&fl--a 29 fRi Ai fR~fR a.n.iu::1.,:~h:1!)ilii)hi::::::ri.Hli' of g. QQ§g 1:;.::gpJg to 0. 02g. 30 31 1~:1:=[]1J~H:::j.:1:1111m111:::::::111~u:1::::::n11i:]:!1i1:!1:~:101:i,1111r.1:::::1111s1:1!::::::1i::: 111£:i:!:1 1 1 1 111 1 32 E.1tr:::::111::1,:J::m::::::1.n~u:::1111v1:tiiii.J:11.m,n1:§,:~:t:1.:j:t:.oiu.t:::::::1::n,;,i::p,w1,u=:t:111n"":,:,:=:111~1r1 33 m1:1:n1iij'ino.1:i1.:,19.:1:::::::o,1:1rY:ii:i.::~:* 34 7

NUMARCQmments March 18, 1993 Une In/Un* Out 1 5. INSTRUMENTATION INSTALLATION 2 3 Ll The instrumentation should be designed and installed *so that tile 4 ¥ihratoFy tFaAsmissihility o¥er the amplifie~ regioA of the ~esigA spectral 5 fFe~YeAcy FaAge is essentially YAity, that is, so that the mounting is rigid. 6 7 Ll The instrumentation should be oriented so that the horizontal axes 8 are parallel to the orthogonal horizontal axes assumed in the seismic 9 an a1Ys i s

• ifl!IlJ1il'!lliJ]i!!l@liIIll!fil:t:tl!ll!!ll!till!!l:!Iil!iii'll!lffl!lllllf ea 11 5.3 . Protection against accidental impacts should be provided.

12 13 6. INSTRUMENTATION ACTUATION 14 15 6.1 Both vertical and horizontal input vibratory ground motion should 16 actuate the same ~ime-history accelerograph. One or more seismic triggers may 17 be used to accomplish this. 18 .1 19 6.2 Spurious triggering should be avoided. 20 6.3 The seismic trigger mechanisms of the time-history accelerograph 22 should be set for a threshold ground acceleration of not more than 0.02g. 23 24 7. REMOTE INDICATION 25 26 Activation of the free-field or any foundation-level time-history 27 accelerograph should be annunciated in the control room. If there are two or 28 more control rooms at the site, annunciation should be provided to each 29 control room. 30 31 8. MAINTENANCE 32 33 8.1 The purpose of the maintenance program is to ensure that the 34 equipment will perform as required. As stated in Regulatory Position 3, the 8

NUMARC c.omments March 18, 1993 Line In/Line Out 1 maintenance and repair procedures should provide for keeping the maximum 2 number of instruments in service during plant operation and shutdown. 3 4 Ll Systems are to be given channel checks every 2 weeks for the first 3 5 months of service after startup. Failures of devices normally occur during 6 initial operation. After the initial 3-month period and 3 consecutive 7 successful checks, monthly channel checks are sufficient. The monthly channel 8 check is to include checking the batteries. The channel functional test 11 12 should be performed every 6 months . Channel calibration should be performed during refueling .

0. IMPLEMENTATION 13 14 The purpose of this section is to provide guidance to applicants and 15 licensees regarding the NRC staff's plans for using this regulatory guide.

16 This proposed revision has been released to encourage public 17 participation in its development . Except in those cases in which the 18 applicant proposes an acceptable alternative method for complying with the 19 specified portions of the Commission's regulations, the method to be described in the active guide reflecting public comments will be used in the evaluation of applications for construction permits, operating licenses, combined 22 licenses, or design certification submitted after the implementation date to 23 be specified in the active guide. This guide would not be used in the 24 evaluation of an application for an operating license submitted after the 25 implementation date to be specified in the active guide if the construction 26 permit was issued prior to that date. 27 28 29 30 31 32 33 34 9

NUMARCC,omments March 18, 1993 Line In/Line Out 1 APPENDIX 2 DEFINITIONS 3 4 Acceleration Sensor. An instrument capable of sensing absolute acceleration 5 and transmitting the data to a recorder. 6 7 Channel Calibration (Primary Calibration). The determination and adjustment, 8 if required, of an instrument, sensor, or system such that it responds within -~ 11 12 a specific range and accuracy to an acceleration, velocity, or displacement input, as applicable, traceable to the National Institute of Standards and Technology (NIST), or responds to an acceptable physical constant. 13 Channel Check. The qualitative verification of the functional status of the 14 instrument sensor. This check is an "in-situ" test and may be the same as a 15 channel functional test. 16 17 Channel Functional Test (Secondary Calibration). The determination without 18 adjustment that an instrument, sensor, or system responds to a known input, 19 not necessarily traced to the National Institute of Standards and Technology -~ 22 23 (NIST), of such character that it will verify the instrument, sensor, or system is functioning in a manner that can be calibrated. Containment - See Primary Containment and Secondary Containment. 24 25 Operating Basis Earthquake Ground Motion (OBE). The vibratory ground motion 26 feF wh i eh these featYFes ef the AYel ea, pe,1eF pl aAt Aeeessary feF eeAt i AYeE:I 27 epe,atieA witheYt YAE:IYe Fisk te the health aAE:I safety ef the pYhlie will 28 FeR1aiA fYAetienal. The 11alYe ef the QBE is set hy the applieaAt ii:iiili:illil 29 !i:11:=111:in1:11:11tllffl:::Jlnl::::=::1::11a11:1:11m1ni:1:1:td::!Mi:i:wl:11J::lltf'!l]:1s11:11::11r:1:11 30 1tn~m~::111:t:t1:!=:11mi11:umm:J:~u~,1-31 32 Primary Containment. The principal structure of a unit that acts as the 33 barrier, after the fuel cladding and reactor pressure boundary, to control the 34 release of radioactive material . The primary containment includes (1) the 10

NUMARCComments March 18, 1993 Une ln/Une Out 1 containment structure and its access openings, penetrations, and appurte-2 nances, (2) the valves, pipes, closed systems, and other components used to 3 isolate the containment atmosphere from the environment, and (3) those systems 4 or portions of systems that, by their system functions, extend the containment 5 structure boundary (e.g., the connecting steam and feedwater piping) and 6 provide effective isolation. 7 8 Recorder. An instrument capable of simultaneously recording the data versus

• 11 12 time from an acceleration sensor or sensors .

Secondary Containment. The structure surrounding the primary containment that acts as a further barrier to control the release of radioactive material. 13 14 Seismje Isolator. A deviee (for iRstaRee, lamiRated elastomer aRd steel) 15 iRstalled hetweeR the strueture aRe its fouReatieR to reeuee the aeeeleratieR 16 of the isolates strueture, as well as the attaehed e~uipmeRt aRd eempoReRts. 17 18 Seismic Trigger. A device that starts the time-history accelerograph. 19 -~ 22 23 Time-History Accelerograph. An instrument capable of measuring and permanently recording the absolute acceleration versus time. The components of the time-history accelerograph (acceleration sensor, recorder, seismic trigger) may be assembled in a self-contained unit or may be separately 24 located. 25 26 Triaxial. Describes the function of an instrument or group of instruments in 27 three mutually orthogonal directions, one of which is vertical. 28 29 30 31 32 33 34 11

NUMARCQ>mments March 18, 1993 Une ln/Une Out 1 REGULATORY ANALYSIS 2 3 A separate regulatory analysis was not prepared for this regulatory 4 guide. The draft regulatory analysis, "Proposed Revision of 10 CFR Part 100 5 and 10 CFR Part 50," provides the regulatory basis for this guide and examines 6 the costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC,

• as Enclosure 2 to Secy 92-215 .

12

NU~Canments March 18, 1993 Une ln/Une Out Enclosure 10 October 1992 Division 1 Draft DG-1017

Contact:

R. M. Kenneally (301) 492-3893 DRAFT REGULATORY GUIDE DG-1017 PRE-EARTHQUAKE PLANNING AND IMMEDIATE NUCLEAR POWER PLANT OPERATOR POST!EARTHQUAKE ACTIONS I A. INTRODUCTION 2 3 Paragraph IV(a)(4) of Proposed Appendix S, "Earthquake Engineering 4 Criteria for Nuclear Power Plarits," to 10 CFR Part 50, "Domestic Licensing 5 of Production and Utilization Facilities," would require that suitable 6 instrumentation 1 be provided so that the seismic response of nuclear power 7 plant features important to safety can be evaluated promptly. Paragraph e: 10 IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 would require shutdown of the nuclear power plant if vibratory ground motion exceeding that of the operating basis earthquake ground motion (OBE) or significant plant damage 11 occurs. Proposed Paragraph 50.54(ee) to 10 CFR Part 50 would require 12 licensees of nuclear power plants that have adopted the earthquake 13 engineering criteria in Proposed Appendix S to 10 CFR Part 50 to shut down 14 the plant if the criteria in Paragraph IV(a)(3) of Proposed Appendix Sare 15 exceeded. 16 This guide is being developed to provide guidance acceptable to the 17 NRC staff for a timely evaluation after an earthquake of the recorded 1 18 Guidance is being developed in Draft Regulatory Guide DG-1016, the Second 19 Proposed Revision 2 to Regulatory Guide 1.12, "Nuclear Power Plant Instru-20 mentation for Earthquakes," to describe seismic instrumentation acceptable to 21 the NRC staff. 22 23 1

NUMARCC,ommenta March 18, 1993 Une ln/lJne Out Enclosure 11 October 1992 Dhision 1 Draft DG-1018

Contact:

R. M. Kenneally (301) 492-3893 DRAFT REGULATORY GUIDE DG-1O18 RESTART OF A NUCLEAR POWER PLANT SHUT DOWN BY A SEISMIC EVENT There are no conments on Draft Regulatory Gu;de DG-1O18. October 1992 Division 1 Draft DG-4003

Contact:

L. Soffer (301)492-3916 DRAFT REGULATORY GUIDE DG-4OO3 (Proposed Revision 2 to Regulatory Guide 4.7) GENERAL SITE SUITABILITY CRITERIA FOR NUCLEAR POWER STATIONS There are no line in/line out conments on Draft Regulatory Guide DG-4OO3. However, general c011111ents regarding the need to update this guidance are given in Enclosures 1 and 2 of this c011111ent package. 1.

N~Comments March 18, 1993 Line In/Line Out Enclosure 12 1 STANDARD REVIEW PLAN 2.5.2 November 1992 2 VIBRATORY GROUND MOTION

Contact:

3 PROPOSED REVISION 3 A. J. Murphy 4 (301)492-3860 5 REVIEW RESPONSIBILITIES 6 Primary - Structural and Geosciences Branch (ESGB) 9, Secondary - None 8 AREAS OF REVIEW 9 The Structural and Geosciences Branch review covers the 10 11 0

   ~=~~~~~ ~~~a~!                             !::abi~~~ 0
                                                          ~!y~;~!1:
                                                                  , 9illlllll1£l,1'~c!~~::!~r:~i~~:

12 the eQafe sil,hutdown e)jarthquake 9Qro'u1i'c! aiotion (SSE) and :the 13 epera:t:1ng basis ear:thqtia.JEe (OBE) for the sife. 'Phe safe sh:u:tdewn 14 ear:thq:uake is :that: ear:thq:uaJEe :that: is based :upen an erval:ua:tien ef 15 :the maxim'l:lHI ear:thq:uaJEe pe:ten:tial censidering :the regienal and leeal 16 geelegy and seismelegy and specific charae:teris:tics ef leeal 17 s\ibs:urfaee ma:terial. I:t is :that: eartilq:uaJEe :that: pred'l:lees tile 18 maxilB'l:llll wibra:tery gre:und me:tien fer whieh safe:ty rela:ted 19 s:tr:ue:t:ures, sys:tems, and eempenen:ts are designed :te remain 20 f:une:tienal. 'Phe epera:ting basis ear:thq:ua)Ee is :that: ear:thq:uaJEe that:, eensidering :the regienal and lecal geelegy, seismelegy, and speeifie eharae:teris:ties ef leeal s:ubs:urfaee ma:terial, ee:uld 23 reasenably be expee:ted :te affect: :the plant: si:te daring :the epera:t 24 ing life ef :the plant, it is that earthq:uaJce that pres.aces the 25 wibratery greens. metien fer which these feat:ures ef the n:uelear 26 pewer plant necessary fer eentin:uea eperatien withe:u:t :una:ue risk te 27 the health ans. safety ef the p'l:lblie are designed te remain *

• 28 f:unetienal. The SSE represents the potential for earthf.fllake greut14 29 *otion at the site an4 is the vibratory ground motion for which 30 certain structures, systems, and components are designed to remain 31 functional. The SSE is based upon a detailed evaluation of the 32 earthquake potential, taking into account regional and local 33 nd

i.ii.~iiw1a.iua;::::!feri;::::n~cfs, 1oe::isi~1!~i!l i1eWi.a sU:~:~~!!:

34 35 !:Ee'r'I'a'f:::;:;:;:;:;:;:;:;:;:;:;t t is defined u as tlle fre'e'! 'l 'I'e'f '! ' ' ' ' g 'r ound response 36 spectra at the plant si te** *. . .* and is described by horizontal and 37 vertical response spectra corresponding to the expected ground 38 motion at the free-field ground surface *~ a hypethetieal ~eek 2.s.2-1

NUMARC C,omments March 18, 1993 Une ln/Une Out 1 euterep. 2 Guidance is lteiag liUlllliin developed on seism~logical and 3 geological investigatlons***"uif. it is described in Draft Regulatory 4 Guide DG-1015, "Identification and Characterization of Seismic iiiis::i1.,1i§i.a!,Waieu:::un~ar:::f::~~'  ;~~sP~IIIIII":!!! 5 6 7 desc::"rlbe*** tbe* * *s*efiimicit*y*******of the site region and the correlation of 8 earthquake activity with seismic sources. seismic sources are 9 10

r.:::E!: 1 11 12 15 16

: ~ ~ ~

any part within -3ff i:IHI Jan (ff& l;i.:i miles) of the site must be identified. More dis"f'ant sources '*'that are capable of earthquakes large enough to affect the site must also be identified. seiamie aeureea eaa lte eapele teeteaie aeurees er aeismegeaie aeureea, aeiameteeteaie pre*.;riaee is a type ef aeismegeaie aeuree. ~ a 17 4etermiaistie an4 preheilistie e?aluatieas are uae4 te assess the 18 BBB. Acceptable 4etermiaistie procedures 1111::::111,:#l~lll,fig'!: : :~fi@J!:au~- 19 are described in this SRP (Subsections 2. s *;*:t:**1****.*th*ro"uglf'* 2*;--5**;--2** :---;1l;: 20 Prelteilistie an4 4etermiaistie methe4s are 4eserilte4 lid in Draft 21 Regulatory Guide DG-1015.  :,: :;: : : : : : : : 22 The principal regulation used by the staff in determining the scope 23 and adequacy of the submitted seismologic and geologic information 24 and attendant procedures and analyses is Appendix A, 11 Seiemie and 25 Geologic Siting Criteria for Nuclear Power Plante II Appendix B, 26 11 Cri teria for the seismic and Geologic Si ting of Nuclear Power 27 Plants on or After [Effective Date of this Regulation)" to 10 CFR Part 100 (Ref. 1). Additional guidance information (regulations, ~30 regulatory guides, and reports) is provided to the staff through References 2 through 8 and 54. 31 Specific areas of review include seismicity (Subsection 2.5.2.1), 32 geologic and tectonic characteristics of the site and region 33 (Subsection 2.5.2.2), correlation of earthquake activity with 34 35 36 0

   ;iiisfiliB:fiii£§t~~~b~~:t i1:n*~~c5~~~~\~

and' 'eent're'i'ii'ng p::;f::: eflll!lff§f!ff! . 9 earthquakes ( Subsection 2

• 5. 2
• 4) , seismic wave 37 transmission characteristics of the site (Subsection 2.5.2.5), and 38 safe shutdown earthquake ground motion (Subsection 2.5.2.6), and 39 operating basis earthquaJce (Subsection 2. 5. 2. 7) .

40 The geotechnical engineering aspects of the site and the models and 41 methods employed in the analysis of soil and foundation response to 42 the ground motion environment are reviewed under SRP Section 2. 5. 4. 43 The results of the geosciences review are used in SRP Sections 44 3.7.1 and 3.7.2. 2.5.2-2

NUMARCC,omments March 18, 1993 Une ln/Une Out 1 II. ACCEPTANCE CRITERIA 2 The applicable regulations (Refs. 1, 2, and 3) a11d regulatory 3 guides (Refs. 4, 5, 6, and 54) and basic acceptance criteria 4 pertinent to the areas of this section of the Standard Review Plan 5 are: 6 1. 10 CFR Part 100, hppentiix A, 11 Seismie anti Geeletjie Si'tinl) 7 Cri:teria fer Nuelear Pewer Plan:ts. 11 Appendix B, "Criteria for 8 the seismic and Geologic Siting of Nuclear Power Plants on or 9 After (Effective Date of this Regulation)

• 11 These criteria 10 describe the Jtintis e! geologic and seismic information needed 11 to determine site suitability and identify geologic and seismic factors required to be taken into account in the siting and design of nuclear power plants (Ref. 1).

14 2. 10 CFR Part 50, Appendix A, "General Design Criteria for 15 Nuclear Power Plants"; General Design Criterion 2, "Design 16 Bases for Protection Against Natural Phenomena" (Ref. 2)

• This 17 criterion requires that safety-related portions of the 18 structures, systems, and components important to safety shall 19 be designed to withstand the effects of earthquakes, tsunamis, 20 and seiches without loss of capability to perform their safety 21 functions.

22 3. 10 CFR Part 100, "Reactor* site Criteria" (Ref. 3). This part 23 describes criteria that guide the evaluation of the 24 suitability of proposed sites for nuclear power and testing 25 reactors. 27 28 29 30

4. Regulatory Guide 1.132, "Site Investigations for Foundations of Nuclear Power Plants." This guide describes programs of site investigations related to geotechnical aspects that would normally meet the needs for evaluating the safety of the site from the standpoint of the performance of foundations under 31 anticipated loading conditions, including an earthquake. It 32 provides general guidance and recommendations for developing 33 site-specific investigation programs as well as specific 34 guidance for conducting subsurface investigations, including 35 the spacing and depth of borings as well as sampling intervals 36 (Ref. 4).

37 5. Regulatory Guide 4.7 (Proposed Revision 2, DG-4003), "General 38 Site Suitability Criteria for Nuclear Power Stations." This 39 guide discusses the major site characteristics related to 40 public health and safety which that the NRC staff considers in 41 determining the suitability of sites for nuclear power 42 stations (Ref. 5). 2.5.2-3

N~Comments March 18, 1993 Une ln/Une Out 1 6. Regulatory Guide 1.60, "Design Response Spectra for Seismic 2 Design of Nuclear Power Plants." ':Ph.is ~ide gi¥es ene met:hed 3 aeeept:able t:e t:h.e NRO st:aff fer defining t:h.e respense speet:ra 4 eerrespending t:e t:h.e eKpeet:ed maximt:HB greund

• aeeelerat:ien 5 (Ref. 6). See alee soothed response spectra are generally 6 used for design purposes - for example, a standard spectral 7 shape that bas been used in the past is presented in 8 Regulatory Guide 1.60 (Ref. 6). These aoothed spectra are 9 **ill acceptable when an appropriate _p~~~ acceleration is used 10 as the high-frequency aeyap,et:e J.Qipfi§~ and the smoothed 11 spectra compare favorably with aite.,;;--s*p*e*c r tic response spectra 12 derived from the Ill_ 9reuall ae,iea estimation procedures discussed in SubsecftTon 2. 5. 2. 6.

-4 13

7. Draft Regulatory Guide DG-1015 (Ref. 54), "Identification and 15 Characterization of seismic sources, Dete:rministie seuree 16 17 18

r-:::ion~r:.d Pl11!it*~11111 1 1111:,,1,1,~111*~~1*1~11,11 prehahilistie an.S dete:rminislle****1 e§jp~i§Ji:!: methodologies for 19 determining the Ill eentrelling e'ar"thquakes for nuclear power 20 plant sites. *************

21 The primary required investigations are described in 10 CFR Part 22 100, in Section IV(a) of Appendix A B(Ref. 1); 'fh.e aeeer,t:able 23 procedures for det:ermining assessing the seismic design bases are 24 given in Sections V(a), (b), and (c). and Seet:ien "II (a) ef t:he 25 ar,r,endix. Draft Regulatory Guide DG-1015 (Ref. 54) is Jteiag 26 lle¥eleped ** providej more detailed guidance on investigations. 27 The seismic design - bases are predicated on a reasonable, conservative determination of the SSE and t:h.e OBB. As defined 4t 31 stated in Sections1/41/41/4 IV and v of Appendix A B(Ref. 1) to 10 CFR Part 100, the SSE and OBB are is based on consideration of the regional and local geology and seismology and on the 32 characteristics of the subsurface materials at the site. aflEl The 33 SSE ~ is described in terms of the vibratory ground motion 34 8 35 =~:::::.v!h.aiJ;ts.i:'*'ffi!t *~:3e./ifidu;:l:; c~~e b~i ter"omu~ a;::"~=h.=~~1~e

• 36 the investic;Ja'i'o n~1;:;: :Afea'ecl" ' ' ' to establish th: &$,@ gseismie g desig~

37 bases; the requirements vary from site to site--:* * * --**----* 38 2.5.2.1 Seismicity. In meeting the requirement of Reference 39 1, this subsection is accepted when the complete historical record 40 of earthquakes in the region is listed and when all available 41 parameters are given for each earthquake in the historical record. 42 The listing should include all earthquakes having Modified Mercalli 43 Intensity (MMI) greater than or equal to IV or magnitude greater 44 than or equal to 3. O that have been reported in all t:eet:enie 45 pre¥inees for all seismic sources, any parts of which are within 2.5.2-4

NUMAFCC,omments March 18, 1993 Line ~/Line Out 1 ff-& i:oo. km ( ~ WIii miles) of the site. A regional-scale map 2 shoufif'he presentecl'''showing all listed earthquake epicenters and 3 should be supplemented by a larger-scale map showing earthquake 4 epicenters of all known events within ff Ml DI (5& I!§ miles) of the 5 site. The following information concerriing eacir**. . . earthquake is 6 required whenever it is available: epicenter coordinates, depth of 7 focus, origin time, highest intensity, magnitude, seismic moment, 8 source mechanism, source dimensions, distance from the site, and 9 any strong-motion recordings (sources from which the information 10 11 =~~h o:staine: s~ou~d s~~ut~e~!if:::~itie~ 1 iHiiia.1ii4:~baiiiii+/-i~Wfi~ In addi :r~n, L 'any' ;~ported earthquake-indu'cea===:=:=:g e'o\ocJfd ' 'f 'a t'i'u'rEt=====11 12 13 llli\l:iiia, such as liquefaction, landsliding' landspreading' arid lti'Fchihg' should be described completely, including the level of strong motion that induced failure and the physical properties of 16 the materials. The completeness of the earthquake history of the 17 region is determined by comparison to published sources of informa-18 tion (e.g., Refs. 9 through 13). When conflicting descriptions of 19 individual earthquakes are found in the published references, the 20 staff should deterinine which is appropriate for licensing 21 decisions. 22 2.5.2.2 Geologic and Tectonic Characteristics of Site and 23 Region. In meeting the requirements of References 1, 2, and 3, 24 this subsection is accepted when all geelegie str\:let\:lres within the 25 regien and teetenie activity seismic sources that are significant 26 in determining the earthquake potential of the region are 27 identified, or when an adequate investigation has been carried out 28 to provide reasonable assurance that all significant teetenie str\:let\:lres seismic sources have been identified . Information presented in Section 2

• 5. 1 of the applicant's safety analysis 31 report (SAR) and information from other sources (e.g., Refs. 9 and 32 14 through 18) dealing with the current tectonic regime should be 33 developed into a coherent, well-documented discussion to be used as 34 ~P:~ }J.c1.s;~s; f'~:r c,1:taraeteri2ing the ear.~.t,;gt1c.\J~e generating petential

                                                                                           ~:~~~

35 36 l!!!! i~ ll~!!!!!!! !! ~Yta~fia::!:::::n::e1!!!!8llr\:l:::~~=s s;::if

• 37 each teetenie previnee seismic source, any part of which is within 38 ff-& i#.jj* DI ( ~ 11:i miles) of the site, must be identified. IJlhe 39 statl~t'nterprets ~a'elseteetenie previnees =te ee regiens ef \:lniferm 40 earth:EJ'-,lcl)~f.! pt3~E.l~~~al._ (s_f.!Js1netE.1etenie pre*vinees) ae_ia ieity (same 41 42  :::ti,,~IJ!::'=lfffP;\~,if~ffJI~\~ t:~d a!!:<<:::::~, e!rer:_eur~~neee~

43 proposed seiseteetenie provinces ii;tl~it.iJP:i!]i.'f:gµ!f§@ may be based oh 44 seismicity studies, differences in "'eicil'o** *. Tct'*hrsffcir , differences in 45 the current tectonic regime, ffe 1i:1:11u1\:i lle!sn:i lis!iiw,ii!ilie.ll'. - 11 46 The staff considers that the mosft .....impcirtari't' ... facb:,*:rs "fcir.......the 47 determination of aeiseteetenie prer+*ii,ees 11:~11~1:t::IB!lris! include 2.5.2-5

NUMARC Comments March 18, 1993 Une ln/Une Out 1 both (1) development and characteristics of the current tectonic 2 regime of the region that is most likely reflected in t.he 3 neeteet:enies (Pest. Mieeene er aee\tt. s in th* Quaternary 4 ( approziately th* last 2 million years and younger geologic 5 history) and (2) the pattern a nd level of historical seismicity. 6 Those characteristics of geologic structure, tectonic history, 7 present and past stress regimes, and seismicity that distinguish the various $,,., :i il[fflciffR~ i i **iaet.eet:enie reYinees and the

   =~:!~~sar!,:~s:h~~i"::" Jlli..rereA~!::~!i:!

8 9 10 11 regienal t.eet:enie medels deri*..-,ed frem aYailable lit.erature seurees, 12 ineluding preYieus SARs and NRO at:aff Safet:y B9aluatien Reperts 13 (SBRs), sheuld be diseussed. ~he medel t:hat: best: eenferms te the tt 16 ebser¥ed dat:a is aeeepted. In addition, in those a r eas where there are capable l.iul:li teeteaie souroes , the results of the additional investigativ*e******r efrjiiirements described in 18 CFR Part 188, Appendix 17 A, Section IV(a) (O) (Ref. 1), SRP Secti on 2 . s . 1 must be presented. 18 ~he diseussien should be augmented by a regienal scale map shewing 19 the teetenie prer.Tinees aeisio aouroes, earthfltta)ce epieent:ers, 20 leeatiens ef geologic structures and ether features t:hat 21 eharact:eri21e t:he aeiso9eaie aouroes (iaeludia9 aeisoteetoaie 22 pro9inees) , and t:he locations ef any eapable fault:s teeteaie 23 sourees. 24 2.5.2.3 Correlation of Earthquake Activity with Geologic Structure 25 Seismegenie Sei]riii6 sources Cinelutling seismeteetenie Pre...,.inees) 26 and eapaele Tectoiilc seurees s£rucHiures er 'Pect01~ic Provinces . In 27 meeting the requirements ot =*= 'tie'feFehce 1, acceptance of this 28 subsection is based on the development of the relationship between the history of earthquake activity and the geelegie structures er aeiaet:eetenic preYinces seismic sources of a region. 4lhe 31 applicant:' s present:at:ien is accepted when t:he earthEfl:laJces diseussed 32 in Subsect:ien 2.5.2.1 ef the SAR are shewn t:e be asseeiated wit:h 33 either geologic structure er tectenie preYince eapele teeteaie 34 11eureea er aeiao9eaie aoureea. Whenever an earthquake hypocenter 35 or concentration of ea;:-_~:11.9:':1~~~- hypocenters can be reasonably . 36 correlated with geelegie #.4§t§ti.¢. structures, the rationale for the 37 association should be devel o'pe.d considering the characteristics of 38 the geologic structure (including geologic and geophysical data, 39 seismicity, and the tectonic history) and the regional tectonic 40 model. The discussion should include identification of the methods 41 used to locate the earthquake hypocenters7 Hd an estimation of 42 their accuracy¥: , and a detailed aceeun:t'*-.w*** t :hat cempares and 43 eentrasts the ""** g eologic struet:ure iw+Telr.Ted in the eart:hf:fl:lalte 44 acti*+Tity with ether areas wi>thin the tectenic preYinee 45 aeiaoteotoaie pre?inee. Particular attention should be given to 46 determining the capability recency and level of activity of faults 47 with which instrumentally located earthquake hypocenters ~ Iii/Ii *.*,***** ................... 2.5.2-6

NUMARC Comments March 18, 1993 Line In/Line Out 1 associated. 2 The presentation should ee augmen=ted ey iDllD.i regio~al maps, all 3 of the same scale, showing the tee=tenie p*r~ine*es seismic sources, 4 the earthquake epicenters, and the locations of geelegie fil.i#.in:£q 5 structures and measurements used to define seismic <<*=*aourcii 6 previnees. Acceptance of the proposed =teetenie previnees seismic 7 sources is based on the staff's independent review of the geologic 8 and seismic information. 9 10 2.s.2.4 Maximum Earthquake Potential Barthpake (OB). In meeting the requirements of Reference 1, this n* een~rellina 11 subsection is accepted when the vibrci:t_e>_i::y gr:-o\ln.cl . :rn<:>:t_i.<:>:ri _due =te from ~  ;~~hm~~~g:!f!!~;e s~:::~::Jt~:S:h!~!~i!!~*~,f!l,f~ :::~~;:;:: 14 asseeia=ted wi=th eaeh =tee=tenie prer.*inee seismic source has been 15 assessed and when the earthquake(&) that would preduee §@I.Iii£ the 16 maximum est se"+"ere vibratory ground motion at the sit*e****** h a*e r** *been 17 determined. The maximum erediele ear=thqualte DSB is =the larges=t 18 ear=th.q ualte tha=t ean reasenaely ee expeeted =te eeeur on a geelogie 19 structure gi"+'"en seismie seuree in the current tectonic regime. 20 eensi4erule ju4gment is in?el?e4 in estiating the agnitu4e of 21 the DSB. iiPiil;il!]iiififi§lilii:~: :, ;,:::l41115.inil ~1 :rncfii R r;ecommended 22 procedures

• 1or**- e stlaitlni;f the** DSB * *are** . aesciribeci *t1i"*braf t
• Regulatory 23 Guide DG-1015 (Ref. 54). Geelegie er seismelegieal evidenee may 24 warrant a maximum earthqualte larger =than =the maxim\::Hll his=torie 25 ear=thqualte. Bar=thqualtes associated with eaeh geolegie strueture er 26 tee=tenic previnee seisie aeuree must ee identified. Where If an 27 earthqualte is asseeiated wi=th a geelegie structure, the maximum erediele earthqualte DSB that could eceur en that strueture should C

30 ee er.ialua=ted, taJEing into aceeunt significant faeters, fer example,

  =the type ef =the faulting, fault length, fault slip rate, rupture 31 length, rupture area, moment, and earthqualte history (e.g. , Refs.

32 19 through 22), 33 In order to determine the maximum erediele earthqualte DSB ilOiJ.Jiff.ffi . 34 IB.lidie.l\lfiiUlilinii that could occur on those faults tiiat*.w.. *are 35 stiown'""ci:f"'"'"'"as'sumed'"'"'"'t'crbe capable teetenie seurees, the staff accepts 36 conservative values based on historic experience in the region and 37 specific considerations of the earthquake history and geologic 38 history of movement on the faults. Where the earthqualtes are 39 aseeeiated wi'th a aeisogenie sou:ree teetenie prer.iince, the larges't 40 histerie earthqualte within t:he aouree province sheuld ee 41 identified. Ieeseismal maps sheuld alee ee presen=ted fer the meet 42 signifieant ear=thqualtes. The II.I. ground metien at the site should 43 be evaluated assuming approp'rTate seismic energy transmission iiliillllllllllll1iiailr:!~~c=d ;~~h ~~~tm;:e~:;::~:i~t= 44 45 2.5.2-7

NUMAAC Comments March 18, 1993 Une ln/Une Out 1 er with eaeh teeteAie pre",..ifiee seismic source:;: eeears at t:he peint 2 ef elesest appreaeh ef the structure er pre,.i1Aee te the site Iii 3 4 ,,w,11:111,,111,1!I11'1ftflNflll\111!:1111111~~::::1~~~:liliml!ill:J:: 1 1 5 The earthquake(s) that would produce the most severe v i b r a t ory

    ~::oa:!11~1~1::if~;:!~~

6 7 8 9 these earthquakes should be specified. The description of the 10 potential earthquake(s) is to 11 include the maxill't\:llll intensity er 18*-il magnitude and ~ 12 IBI.IAW.I distance from the assl:iJftea :,:,:*:*:1e:*ea'Ten . ef the pe:tential tt! eeirli'liijiialte(s) te the site. Per the seisaeteett'Htie pre?iaee surreuadiag the site, the DSB is assumed te eeeur eeut 15 Jea frea 15 16 t~~d~~:~* ;;he P Y :~:ff1 !~d:;:nd:~;!~ EJ ~:~t~ua= EJr t?,.~:~i.itlei'ITT't~i2~,i&i. U.Jt. . ,.,. .,.,.,. .,.,.,. .,.,. .,.,.,.,.,.,.,.,.,.,.,. .,.,. . .,. . . .,.,. .,.,. . Jl\\.. .*.************' 17 associated with each geelegie struetare er**************fiieti'en'Ie*************ji*r*ev'i'iie'e 18 seismic source.

• 19 Controll~J:lg .~a.~:tl:l'l'l.~~es are those earthquakes that preduee the 20 largest R#l#ll.Uli]Ji.6 ground motion at the nuclear power plant 21 site. Pr*ocedti'reiis" .....for deriving controlling earthquakes from a 22 probabilistic seismic hazard analysis are discussed in Appendix c 23 of Draft Regulatory Guide DG-1015 (Ref. 54). Acceptance of the 24 description of the peteAtial controlling earthquakes that weula 25 preaaee the largest greana metien at the sH:e is based on the 26 staff's independent analysis.

-728 29 2.5.2.5 Seismic Wave Transmission Characteristics of the Site. In meeting the requirements of Reference 1, this subsection is accepted when the seismic wave transmission characteristics 30 (amplification or deamplification) of the materials overlying 31 bedrock at the site are described as a function of the significant 32 frequencies IfilRl:1:Mill!i:J:!. The following material properties should be . 33 determined fc:i'r*******ea*cfi'*****i;tratum under the site: seismic compressional 34 and shear wave velocities, bulk densities, soil index properties 35 and classification, 1t1111ililllii shear modulus and damping 36 :r.iariatiens with strafn*******Terier;************a'rid*********water table elevation ana its 37 wariatien. In each case, methods used to determine the properties 38 should be described in Subsection 2. 5. 4 of the SAR and cross-39 referenced in this subsection. For the maximum earthE!l:!alte CE ( s ) 40 determined in Subsection 2. 5. 2. 4 and Draft Regulatory Guide DG-1015 41 (Ref. 54) , the free-field ground motion ( including significant 42 frequencies) must be determined, and an analysis sheala ee 43 performed to determine the site effects on aifferefit seismic wave 44 types imli:1111!:I in the significant frequency bands  ::,111!:t::::tl!!~::. H 2.5.2-8

NUMARCQmments March 18, 1993 Line In/Line Out 1 appropriate, the analysis sheulti eensitier the effects ef sit:e 2 eentiitiens anti material preperty ?ariatiens upen wa?e prepagatien 3 anti fref!Ueney eentent. 4 The free-field fl.$! greunti metien (also referred to as control 5 motion) should l>e=*==*'aefined t:e be en a itiH\tiHi ground surface:¥ eM 6 sheulti be based *e n tiata ebtained in the~ ' 'f're'e ' 'f ield. 'l?we ease's are 7 identified, depending en the seil eharacteristics at the site anti 8 subject te availability ef apprepriate recertied greund metien tiata. 9 When tiata are available, fer example, fer relati?ely uniferm sites 10 ef seil er reek with smeeth ?ariatien ef properties with depth, the 11 eentrel peint (lecatien at which the eentrel metien is applied) -~ 12 15 16 17 sheuld be specified en the seil surface at the tep ef the ai~* finished grade. The free-field ~SE greund met.ien or control motion should be consistent with the properties of the soil profile. Fer sites eempeaeti ef ene er mere thin seil layers e?erlying a eempet.ent material, er in t;!le ease ef insufficient reeerdeti greunti met.ien data, the cent.rel paint is specified en an eut.erep er a 18 hypet.hetieal eutcrep at a leeat.ien en the tep ef the e~mpetent. 19 material. ~he cent.rel metien speeifieti aheulti be eensistent. with 20 the prepert.ies ef the cempetent material. 21 Where vertically propagating shear waves may produce the maximum 22 ground motion, a one-dimensional equivalent-linear analysis (e.g., taa~~t!!,M~~ 23 24 25 26 27 30 31 32 maximum greund metien, et.her met.beds ef analysis (e.g., Refs. 28 anti 29) may be mere apprepriat.e. Hewer+"er, since seme ef the

   ?ariables are net well defined anti the t.eehnif!Ues are at.ill in the tier+"elepmental st.age, ne generally agreed upen precetiurea can be 33 premulgat.eti at this time. Hence' t. The staff mus-t. wilI use discre-34 tion in reviewing any method of analysis~*=*=*=*:*=*=*=*=*=* == To ensure 35 appropriateness, site 36 response characteristics determined from analytical procedures 37 should be compared with historical and instrumental earthquake 38 data, when available.

39 2. 5. 2. 6 Safe Shutdown Earthquake Ground Motion ~:1:ss11:: . 40 In meeting the requ i rements of Reference 1, th i s subsectTon is 41 accepted when the ?ibratory ground motion specified for the SSE is 42 described in terms of the free-field response spectrum and is at 43 least as conservative as that which would result at the site from 44 the maximum eart.hf!Ualfe CEs determined in Subsection 2

• 5. 2
• 4, 45 considering the site !!iffij§ffl1/4:Bfli!X@ transmission effects determined 2.5.2-9

NUMARCQ>mments March 18, 1993 Une ln/Une Out 1 in Subsection 2. 5. 2. 5. If several different :maxilMHll petential 2 earthf.!'Uakes CBs produce the largest ground motions in different 3 frequency bands (as noted in Subsection 2.5.2.4), ~he vibratory 4 ground motion specified for the SSE must be as conservative in each 5 6 uii:iiibiiiii:B~:i:Bii1:m- 111:::::::::eltilllllw:lnl earthquake  ::~:111 7 The staff reviews the free-field response spectra of engineering 8 significance (at appropriate damping values). Ground motion may 9 vary for different foundation conditions at the site. When the 10 site effects are significant, this review is made in conjunction 11 12 15 16 with the review of the design response spectra in Section 3.7.1 to i&i~i11: Tu1$.:i:;:i:i11i:: lia~1u;:i1iiliii:\:i~i1.l0s,m11IIIIIIIIIIIIIBIII i itl@w.#.ilf.J::11,a§i)IIMI[ggijij;i)iiii1b.§\Illiiv.ifiel}Jl@iln4.:ln1Ifjµ§[t)\i/1:\iliY][}(§# ffilffl!siPitll]J]iYl!@ffii::t: : : :111:: : 2emti2nlfil:1:: :11ii1m11s:: :1eri~:1:1111 ~r-***,The**

• s tirt r 1

iiormalry"***evaiuat es****re*spo"iise** *s pect ra*** on****i:i"****aase*.:;;by".:;;*c as*e ****has is . The 17 staff considers compliance with the following conditions acceptable 18 in the evaluation of the SSE. In all these procedures, the 19 proposed free-field response spectra shall will be considered 20 acceptable if they equal or exceed the estimated 84th percentile 21 ground-motion spectra ii.ii@ from the :maxim\:Hft er eentrelling 22 earthf.!tia]f:e CE described *lii"**subsection 2. 5. 2. 4. 23 The following steps summarize the staff review of the SSE. 24 1. Both horizontal and vertical component site-specific response 25 spectra should be developed statistieally frem respense -~ 26 29 30 speetra ef reeerded streng metien reeerds that are seleeted te ha*+"'e similar seuree, prepagatien path, and reeerding site preperties as the eentrelling earthf.!tia]f:es. It :must be ensured

           ;;;;nii:~iiiiti&i§~i~ii;.;ni:i ii~iq\l.iRin~::;:~~~:: llfllll.fIll 1

31 32 33 34 11!11,~~~!llilllt?!!?e1!1!!r:!!. be present beeause ef leeatiens aneljer heusing ef reeerding 35 instruments. I:mpertant seuree preperties inelude :magnitude 36 and, if pessible, fault type, and t:eetenie ew+"'irenment. 37 Prepagatien path preperties inelude distanee, depth, and 38 attenuatien. Relevant site properties include shear velocity 39 profile and other factors that affect the amplitude of waves 40 at different frequencies. A suffieiently large nu:ml,er ef 41 site speeifie time histeries er respense speetra er beth 42 sheuld be used te ebta:in an adef.!tiately breadbana speetru:m te 43 eneempass the uneertainties in these parameters. An 84th per-44 centile response spectrum fer the reeerds should be presented 45 for each damping value of interest and compared to the SSE 2.5.2-10

NUMARC Comments March 18, 1993 Une ln/Une Out 1 free-field and design response spectrum (e.~., Refs. 38, 31, 2 32, anci 33)

• The staff considers direct estimates of spectral 3 ordinates preferable to scaling of spectra to peak 4 accelerations. In the Eastern United States*, relatively 5 little information is available on magnitudes for the larger 6 historic earthquakes: hence, it may be apprepriate te rely en 7 int:ensi:ty ebservat:iens (cieseript:iens ef earthtftia]ce effeets) te 8 est:imat:e magni:tecies ef hist:erie ewent:s (e,,,, Refs. 34 anci 9 35). If t:he cia:ta fer site apeeifie respense speetra were net 10 ebtained encier geelegie eenciitiens similar te these at the 11 sit:e, eerreetiens fer site effeet:s sheelci be inelecieci in the liiiiiiiii;iEi§nli11fil:1ii1iiii*iiiii:iiiiiii~i:1l~i:l:iil::I:::::B 12

.4 13 15 16

2. Where a large eneegh enseele ef st:reng met:ien reeercis is net available, respense speetra may be appreKimat:ed by sealing t:hat ensemble ef st:reng metien ciat:a t:hat: represent: t:he best:

17 estimate ef seeree, prepagatien path, anci site J>reperties 18 (e.g., Ref. 36). Sensitivity steelies sheelci shew t:he effects 19 ef sealifl<J. 20 3. If streng metien reeercis are net available, si:te speeifie J>eak 21 greenci aeeeleratien, r+"eleeity, anci ciisplaeement (if necessary) 22 sheelci be determined fer apprepriat:e magnitecie, dist:anee, anci 23 feendatien eenditiens. ~hen resJ)ense speet:ra may be 24 determined by sealing the aeeeleratien, veleeity, and 25 disJ>laeement: walees by apprepriat:e amplifieat:ien faeters 26 (e.g., Ref. 37). Where If enly estimat:es ef J>eaJc greenci 27 aeeeleratien are available, it is aeeeptable te seleet a peak aeeeleratien and ese this peaJc aeeeleratien as the high freqeeney asymptete te standarcii2eci respense speetra seeh as 30 deseribeci in Regelatery Seide 1. 68 (Ref. 6) fer beth the 31 heri2ental anci vertieal eempenents ef metien with the 32 apprepriate amplifieatien faeters. Fer eaeh eentrelling 33 earthqeaJee, the peaJe greend metiens sheeld be determined esing 34 eerrene relat:iens between aeeeleratien, rveleeiey, and, if 35 neeessaey, displaeement:, ea!.'thtftia)ee si2e (magnitecie er 36 intensity) , and seeree diseanee. PeaJe greend metien sheeld be 37 determined frem state ef the art relatienships. Relatienships 38 between magnitecie and greend metien are feend, fer eKample, in 39 Referenees 3 8, 3 9, 4 8, and 41 and relatienshiJ>s bet:ween greend 40 metien and int:ensity are feend, fer eKample, in References 41, 41 42, and 43. Dee te Beeauae ef the limited data fer high 42 intensities great:er than Medifiea Merealli Intensit:y (MMI) 43 VIII, the available empirieal relat:ienships between int:ensity 44 and peaJe greend met:ien may net be seit:able fer det:ermining t:he 45 apprepriat:e referenee aeeeleratien fer seismie design. 2.s.2-11

NUMARC Comments March 18, 1993 Une ln/Une Out 1 2 3 4

           ~:;!ti:~~=:?t:;:~i:;:e:~~::~~

the appropriateness of the model are thoroughly documented 5 (e.g., Refs. 19, 44, 45, and 46, eft4 53, ih.lD]Ht). Modeling is 6 particularly useful for sites near capabt*e*=*=*=*=*i aults teeteaie 7 aeureea or for deeper structures that may experience ground 8 motion that is different in terms of frequency content and 9 wave type from ground motion caused by more distant 10 earthquakes. 11 s. Preeaeilistic estimates ef seismic haeard shettld ee calcttlated 12 (e.g., Refs. 41 and 47) and the ttnderlying assttmptiens and ~ asseciated ttncertainties shettld ee dec\:l'lltented as diaeuaaed ia Braft Regulatory Guide BG 1815 (Ref, 54), te assist in the 15 staff's e¥erall deterministic appreach. 'fhe preeaeilistic 16 stttdies shettld highlight which seismic settrces are significant 17 te the site. Uniferm haeard spectra (spectra that ha¥e a 18 ttniferm preeaeility ef eMceedance e¥er the freEjttency range ef 19 interest) shewing ttncertainty shettld ee calcttlated fer 0.01, 20 0.001, and 0.0001 annttal preeaeilities ef eMceedance at the 21 site. 'fhe aaaual preeaeility ef eMceeding the SSE respense 22 spectra shettld alee ee estimated and cemparedisen ef resttlts 23 made with ether preeaeilistic stttdies. witll tlloae of tile 24 eurreatly operating pleats as required ia Appendix B to 18 CPR 25 Part 188, Proeedures for 4eri1+"iag tile CBs fro a PSHA aad 26 eatiatiag tile aaaual prebeility of e*eeedaaee ef tile SSB are 27 eoataiae4 ia tile Braft Regulatory Guide BG 1815 (Appeadieea B 28 aa4 C) (Ref, 54), - The time duration and number of cycles of strong ground motion are 30 required for analysis of site foundation liquefaction potential and 31 for design of many plant components. The adequacy of the time 32 history for structural analysis is reviewed under SRP Section 33 3. 7 .1. The time history is reviewed in this SRP section to confirm 34 that it is compatible with the seismological and geological . 35 conditions in the site vicinity and with the accepted SSE medel. 36 At present, models for deterministically computing the time history 37 of strong ground motion from a given source-site configuration may 38 be limited. It is therefore acceptable to use an ensemble of 39 ground-motion time histories from earthquakes with similar size, 40 site-source characteristics, and spectral characteristics or 41 results of a statistical analysis of such an ensemble. Total 42 duration of the motion is acceptable when it is as conservative as 43 values determined using current studies such as References 48, . 49, 44 50, and 51. 45 2. 5. 2. 7 Operating Basis EarthEftlalce. In meeting the 2.5.2-12

                                                                                          ~ Comments March 18, 1993 Une ln/Une Out 1 requirements ef Referenee 1, this eeseetien is aeeeptaele when the 2 wieratery greand metien fer the OBB is deserieed and the respense 3 speetrtHB (at appre~iate damping walaes) at the site epeeified.

4 Preeaeility ealealatiens (e.g., Refs. 41, 47, and 52) eheald be 5 ased te estimate the preeaeility ef eMeeeding the OBB daring the 6 eperating life ef the plant. 'Phe maMi!ll:Hll wieratery greand metien 7 ef the OBS eheald be at least ene half the maMimtHll wibratory greand 8 metien ef the 66B anless a lever OBB ean be jastified on the basis 9 ef prebability ealealatiens. It has been staff praetiee te aeeept 10 the OBS if the retarn period is en the order ef handreds ef years 11 (e.g., Ref. 31). .3 12 14 15 III. REVIEW PROCEDURES Upon receiving the applicant's SAR, an acceptance review is conducted to determine compliance with the requirements of 10 CFR Part 100, Appendix A B (Ref. 1)

• investigative The 16 reviewer also identifies any site-specific problems, the resolution 17 of which could result in extended delays in completing the review.

11,,~~ 18 After SAR acceptance and docketing, these areas are identified 19 where the reviewer identifies areas that need additional 20 21 22 iii1ii!ii: i11ii*~iil1;t~ii: riiitted *to

• the******

                                 ~&1iiii:i i:;imi.§i:1iii&:a ;°ii:::a::e            1
                            *applfcarit********a:s*******d"r*a -rt *******r equei;ts for additional 23 information.

24 A site visit may be conducted during which the reviewer inspects 25 the geologic conditions at the site and the region around the site as shown in outcrops, borings, geophysical data, trenches, and ~28 those geologic conditions exposed during construction if the review is for an operating license. The reviewer also discusses the 29 questions with the applicant and his consultants so that it is 30 clearly understood what additional information is required by the 31 staff to continue the review. Following the site visit, a revised 32 set of requests for additional information, including any 33 additional questions that may have been developed during the site

• 34 visit, is formally transmitted to the applicant.

35 The reviewer evaluates the applicant's response to the questions, 36 prepares requests for Bl additional elarifying information, and 37 formulates positions thiit** may agree or disagree with those of the 38 applicant. These are formally transmitted to the applicant. 39 The Safety Analysis Report and amendments responding to the 40 requests for additional information are reviewed to determine that 41 the information presented by the applicant is acceptable according 42 to the criteria described in Section II (Acceptance criteria) 2.5.2-13

NUMAACC,omments March 18, 1993 Une ln/Une Out 1 above. Based on information supplied by the applicant and 2 information obtained from site visits, er frem staff consultants, 3 or literature sources, the reviewer independently identifies and 4 evaluates the relevant seismeteetenie previnees aeisogenie li~i.DI 5 sources and capaJ>le !i.iiltii ileetlonie aourees, e*.,.alttat,i"s***...*.* h*e 6 eapaeility et fauH:s in*~i.l'e::::::--regien, and determines the earthquake 7 potential for each)f) pre"-vinee and eaeh eapaele faul:e er :eee:eenie 8 strue:eure **iaogeiie aeure* er eapele ileetlenie aeuree using 9 procedures noted in Section II (Acceptance Criteria) above. The 10 reviewer evaluates the vibratory ground motion that the pe:een:eial 11 earthffl:la)ces CEs could produce at the site and defines compares that 12 ground otion to the SSE . safe shutElewfl ear'thffl:la:Jce anEl epera:eing easis ear'thffl:laJce. 4': IV. EVALUATION FINDINGS 15 If the evalua:eien by 'the s:eaff, On completion of the review of the 16 geologic and seismologic aspects of the plant site, if the 17 evaluation by the staff confirms that the applicant has met the 18 requirements or guidance of applicable portions of References 1 19 through 6 and* ff II, the conclusion in the SER states that the 20 information provid.ed and investigations performed support the 21 applicant's conclusions regarding the seismic integrity of the 22 subject nuclear power plant site. In addition to the conclusion, 23 this section of the SER include1:3._ .. PJ ~efini:eiens an evaluati.C?~ .C?f 24 :eee:eenie pre"+*inees aeisegenie iiliillli sources and capaJ>le Iii.I.Ii 25 tleetenie sourees, ( 2) evalua:eieiis**********e*!**. . . 'the eapabil i:ey ef ge*eYe*ijTe 26 s:erue:eures in 'the regien, (-a- Ii) detel'lltina:eiens evaluation of the 27 66:B ear:ehqua:Jte (s) BSBs anEl free-field response spectra based on evaluation of the pe:een:eial ear:ehquaJces CBs, and (+ I) time-history of strong ground motion, and (5) Ele'tel'lltina:eiens et** 'the OBB free 30 field respense speetra. Staff reservations about any significant 31 deficiency presented in the applicant's SAR are stated in 32 sufficient detail to make clear the precise nature of the concern. 33 ~he abeve evaluatiens Eletel'lltinatiens er rede'tel'lltina:eiens are ma:Ele 34 by the s:eaf f during beth the eenstrttetien pel'lltit ( CP) , and . 35 operating license (OL), eoeined lieense (OOL) or early site perit 36 phases ef review as appropriate. 37 OL appl iea:eiens are reviewed fer any new infel'lltatien de"+*eleped 38 subsefft:1en:e :ee 'the CF safe:ey evaluatien reper:e SBR. The I.I review 39 will also determine whether the CP recommendations :ti'a'.°=ve been 40 implemented. 41 A typical OL-stage summary finding for this section of the SER 42 follows: 43 In our review of the seismologic aspects of the plant site, we 2.5.2-14

NUMAFC c.omments March 18, 1993 Une ln/Une Out 1 have considered pertinent information gathered since our 2 initial seismologic review which that was made in conjunction 3 with the issuance of the Construction Permit.. This new 4 information includes data gained from both site and near-site 5 investigations as well as from a review of recently published 6 literature. 7 As a result of our recent review of the seismologic 8 information, we have determined that our earlier conclusions. 9 regarding the safety of the plant from a seismologica'i ' 10 standpoint remain& valid. These conclusions can be summarized 11 as follows: ~; 1. Seismologic information provided by the applicant and required by A~pendix A B to _10 CFR Part 100,,,,J?,,:,g y,,!.~;,~, -, , ,! I} 14 adequate basis to establish that no J.tl.i9:n:¥.lht9.U.~'f 15 inigf.#iUni~a capable faults seismic sources ii5i"fst****--rff. . . .tlie 16 p'lahf'"'sil:e *<*a reaM which tllat weals cause earthqua]tes te 1,e 17 centereei there~. . .. 18 2. The response sp_-a.c::tnHllfi proposed for the safe shutdown 19 earthquake 1/4s If.~ the* appropriate free-field response 20 spectrHI. in coiifc5"rmance with Appendix A B to 10 CFR Part 21 100. *:-:-:-: 22 The new information reviewed for the proposed nuclear power 23 plant is discussed in Safety Evaluation Report Section 2.5.2. -~ 24 27 28 The staff concludes that the site is acceptable from a seismologic standpoint and meets the requirements of (1) 10 CFR Part SO, Appendix A (General Design Criterion 2), (2) 10 CFR Part 100, and (3) 10 CFR Part 100, Appendix A B. This conclusion is based on the following; 29 1. The applicant has met the requirements of: 30 a. 10 CFR Part so, Appendix A, General Design 31 Criterion 2 with respect to protection against 32 natural phenomena such as faulting. 33 b. 10 CFR Part 100, Reactor Site Criteria, with 34 respect to the identification of geologic and 35 seismic information used in determining the 36 suitability of the site. 37 c. 10 CFR Part 100, Append i x A (Seismi c and Geel egie 38 Siting Criteria fer Nttclear Fewer Plants) Appendix 39 B (Criteria for the Seismic and Geologic Si ting of 2.5.2-15

                                                                             ~Comments March 18, 1993 Une ln/Une Out 1                  Nuclear Power Plants on or After [Bffective Date of 2                  this Regulation]                (Ref. 1))  with respect to 3                  obtaining the geologic and seismic. information 4                  necessary to determine (1) site suitability and (2) 5                  the appropriate design of the plant. Guidance for 6                  complying with this regulation is contained in 7                  Regulatory Guide 1.132, "Site Investigations for 8                  Foundations of Nuclear Power Plants" (Ref. 4) ;

9 Draft Regulatory Guide DG-1015, "Identification and 10 Characterization of Seismic Sources, Deterministie 11 12 n~lii.iiur;::~n:n~!ll*~111~*:  := : 1 1::~:

                                                                          ~1r11~~l,~e 13                  Gulcfit**-*~r:**1***;* "General Site Suitability Criteria for

~ Nuclear Power Stations" (Proposed Revision 2) (Ref. 5); and Regulatory Guide 1.60, "Design Response 16 Spectra for Seismic Design of Nuclear Power Plants" 17 (Ref. 6). 18 V. IMPLEMENTATION 19 The following is intended to provide guidance to applicants and 20 licensees regarding the NRC staff's plans for using this SRP 21 section. 22 Except in those cases in which the applicant or licensee proposes 23 an acceptable alternative method for complying with specific 24 portions of the Commission's regulations, the methods described 25 herein will be used by the staff in its evaluation of conformance 26 with Commission regulations. e, 28 Implementation schedules for conformance to parts of the method discussed herein are contained in the referenced regulatory guides 29 and NUREGs (Refs. 4 through 8 and 54). 30 The provisions of this SRP section apply to reviews of construction 31 permits (CP), operating licenses (OL), early site permits, . 32 prelimiHary design appro¥al (PDA), fiHal desigft appre¥al (FDA), and 33 combined license (CP/OL) applications docketed pursuant to the 34 proposed Appendix B to 10 CFR Part 100. after the date of issaaftee 35 of this SRP seetieft. 36 VI. REFERENCES 37 1. 10 CFR Part 100, 1.rppeHdix A, 11 6eismie aftd Geologie Siting 38 Oriteria for Naelear Power Plants. 11 Proposed Appendix B, 39 "Criteria for the Seismic and Geologic Siting of Nuclear Power 40 Plants on or After [Effective Date of this Regulation]." 2.5.2-16

NUMAFC Comments March 18, 1993 Line In/Line Out 1 2* 1 o CFR Part 5 o, Appendix A, General Design Criterion 2 , 2 "Design Bases for Protection Against Natural Phenomena." 3 3. 10 CFR Part 100, "Reactor Site Criteria." 4 4. USNRC, "Site Investigations for Foundations of Nuclear Power 5 Plants," Regulatory Guide 1.132. 6 5. USNRC, "General Site Suitability Criteria for Nuclear Power 7 Stations, " Regulatory Guide 4. 7 ( Proposed Revision 2, DG-8 4003). eo9 6. USNRC, "Design Response Spectra for Seismic Design of Nuclear Power Plants," Regulatory Guide 1.60. 11 7. us NRC, "Standard Format and Content of Safety Analysis 12 Reports for Nuclear Power Plants (LWR Edition)," Regulatory 13 Guide 1.70. 14 8. USNRC, "Report of Siting Policy Task Force," NUREG-0625, 15 August 1979. 16 9. N. L. Barstow et al., "An Approach to Seismic Zonation for 17 Siting Nuclear Electric Power Generating Facilities in the 18 Eastern United States," prepared by Roundout Associates, Inc. , 19 for the USNRC, NUREG/CR-1577, May 1981. 20 10. C. W. Stover et al., "Seismicity Maps of the States of the 21 U. s. , " Geological Survey Miscellaneous Field Studies Maps, 92 1979-1981. 23 11. "Earthquake History of the United States," Publication 41-1, 24 National Oceanic and Atmospheric Administration, U.S. 25 Department of Commerce, 1982. 26 12. T. R. Toppozada, c. R. Real, s. P. Bezore, and D. L. Parke, . 27 "Compilation of Pre-1900 California Earthquake History, Annual 28 Technical Report-Fiscal Year 1978-79, Open File Report 79-6 29 SAC (Abridged Version)," California Division of Mines and 30 Geology, 1979. 31 13. P. w. Basham, o. H. Weichert, and M. J. Berry, "Regional 32 Assessment of Seismic Risk in Eastern Canada," Bulletin of the 33 Seismological Society of America, Vol. 65, pp. 1567-1602, 34 1979. 35 14. P. B. King, "The Tectonics of North America - A Discussion to 36 Accompany the Tectonic Map of North America, Scale 2.5.2-17

NU~Commems March 18, 1993 Line ln/Une Out 1 1:5,000,000," Professional Paper 628, U.S. Geological Survey, 2 1969. 3 15. A. J. Eardley, "Tectonic Divisions of North America," Bulletin 4 of the American Association of Petroleum Geologists, Vol. 35, 5 1951. 6 16. J. B. Hadley and J. F. Devine, "Seismotectonic Map of

• the 7 Eastern United States," Publication MF-620, U.S. Geological 8 Survey, 1974.

9 17. M. L. Sbar and L. R. Sykes, "Contemporary Compressive Stress -~ 10 13 14 18. and Seismicity in Eastern North America: An Example of Intra-Plate Tectonics, " Bulletin of the Geological Society of America, Vol. 84, 1973. R. B. Smith and M. L. Sbar, "Contemporary Tectonics and Seismicity of the Western United States with Emphasis on the 15 Intermountain Seismic Belt," Bulletin of the Geological 16 Society of America, Vol. 85, 1974. 17 19. USNRC, "Safety Evaluation Report (Geology and Seismology) 18 Related to the Operation of San Onofre Nuclear Generating 19 Station, Units 2 and 3," NUREG-0712, February 1981. 20 20. D. B. Slemmons, "Determination of Design Earthquake Magnitudes 21 for Microzonation," Proceedings of the Third International 22 Earthquake Microzonation Conference, 1982.

21. M. G. Bonilla, R. K. Mark, and J. J. Lienkaemper, "Statistical Relations Among Earthquake Magnitude, Surface Rupture, Length 25 and Surface Fault Displacement," Bulletin of the Seismological 26 Society of America, Vol. 74, pp. 2379-2411, 1984.

27 22. T. c. Hanks and H. Kanamori, "A Moment Magnitude Scale," 28 Journal of Geophysical Research, Vol. 84, pp. 2348-2350, 1979 * . 29 23. P. B. Schnabel, J. Lysmer, and H.B. Seed, "SHAKE-A Computer 30 Program for Earthquake Response Analysis of Horizontally 31 Layered Sites," Report No. EERC 72-12, Earthquake Engineering 32 Research Center, University of California, Berkeley, 1972. 33 24. E. Faccioli and J. Ramirez, "Earthquake Response of Nonlinear 34 Hysteretic Soil Systems, " International Journal of Earthquake 35 Engineering and Structural Dynamics, Vol. 4, pp. 261-276, 36 1976. 37 25. I. v. Constantopoulos, "Amplification Studies for a Nonlinear 2.5.2-18

NUMAFC Comments March 18, 1993 Une In/Un* Out 1 Hysteretic Soil Model," Report No. R73-46, Department of Civil 2 Engineering, Massachusetts Institute of Technology, 1973. 3 26. v. L. Streeter, E. B. Wylie, and F. E. Richart, ' 11 soil Motion 4 Computation by Characteristics Methods," Proceedings of the 5 American Society of Civil Engineers , Journal of the 6 Geotechnical Engineering Division , Vol. 100, pp. 247-263, 7 1974. 8 27. w. B. Joyner and A. T. F. Chen, "Calculations of Nonlinear 9 Ground Response in Earthquakes," Bulletin of the Seismological 10 Society of America, Vol. 65, pp. 1315-1336, 1975. 28

• T. Udaka, J. Lysmer, and H. B. Seed, "Dynamic Response of Horizontally Layered Systems Subjected to Traveling Seismic Waves, " Proceedings of the Second U. s. National Conference on Earthquake Engineering . 1979.

15 29. L. A. Drake, "Love and Raleigh Waves in an Irregular Soil 16 Layer," Bulletin of the Seismological Society of America, Vol. 17 70, pp. 571-582, 1980. 18 30. USNRC, "Development of Site-Specific Response Spectra," 19 NUREG/CR-4861, March 1987. 20 31. USNRC, "Safety Evaluation Report Related to Operation of the 21 Sequoyah Nuclear Plant, .Units 1 and 2," NUREG-0011, 1979. 22 32. USNRC, "Safety Evaluation Report Related to the Operation of Midland Plant, Units 1 and 2, 11 NUREG-0793, May 1982 *

• 3 24 33. USNRC, "Safety Evaluation Report Related to the Operation of 25 Enrico Fermi Atomic Power Plant, Unit No. 2," NUREG-084 7, July 26 1981.

27 34

• R. L. Street and F. T. Turcotte, "A study of Northeastern .

28 North American Spectral Moments, Magnitudes, and Intensities, " 29 Bulletin of the Seismological Society of America. Vol. 67, pp. 30 599-614, 1977. 31 35. O. W. Nuttli, G. A. Bollinger, and D. W. Griffiths, "On the 32 Relation Between Modified Mercalli Intensity and Body-Wave 33 Magnitude," Bulletin of the Seismological society of America, 34 Vol. 69, pp. 893-909, 1979. 35 36. T. H. Heaton, F. Tajima, and A. w. Mori, "Estimating Ground 36 Motions Using Recorded Accelerograms," Surveys in Geophysics, 37 Vol. 8, pp. 25-83, 1986. . 2.5.2-19

NUMARCComments March 18, 1993 Line In/Line Out 1 37. USNRC, "Development of Criter1a for Seismic Review of Selected 2 Nuclear Power Plants," NUREG/CR-0098, June 1978. 3 38. w. B. Joyner and o. M. Boore, "Peak Horizontal Acceleration 4 and Velocity from Strong Motion Records Including Records from 5 the 1979 Imperial Valley, California Earthquake," Bulletin of 6 the Seismological Society of America, Vol. 71, 2011-2038, 7 1981. 8 39. K. W. Campbell, "Near-Source Attenuation of Peak Horizontal 9 Acceleration," Bulletin of the Seismological Society of 10 America, Vol. 71, pp. 2039-2070, 1981.

40. o. w. Nuttli and R. B. Herrmann, "Consequences of Earthquakes in the Mississippi Valley," Preprint 81-519, American Society of Civil Engineers Meeting, 1981.

14 41. D. L. Bernreuter et al., "Seismic Hazard Characterization of 15 69 Nuclear Plant Sites .East of the Rocky Mountains," NUREG/CR-16 5250, January 1989. 17 42. M. D. Trifunac and A. G. Brady, "On the Correlation of Seismic 18 Intensity Scales with Peaks of Recorded Strong Ground Motion," 19 Bulletin of the Seismological Society of America, Vol. 65, 20 1975. 21 43. J. R. Murphy and L. J. O'Brien, "Analysis of a Worldwide 22 Strong Motion Data Sample To Develop an Improved Correlation 23 Between Peak Acceleration, Seismic Intensity and Other Physical Parameters," prepared by Computer Sciences Corporation for the USNRC, NUREG-0402, January 1978. 26 44. USNRC, "Safety Evaluation Report Related to Operation of 27 Virgil c. Summer Nuclear Station, Unit No. 1," NUREG-0717, 28 1981. 29 45. USNRC, "State-of-the-Art Study Concerning Near-Field 30 Earthquake Ground Motion," NUREG/CR-1340, August 1980. 31 46. H. J. Swanger et al., "State-of-the-Art study Concerning Near-32 Field Earthquake Ground Motion," NUREG/CR-1978, March 1981. 33 4 7. "Seismic Hazard Methodology for the Central and Eastern United 34 States," Electric Power Research Institute, Report NP-4726, 35 1986. 36 48. R. Dobry, I. M. Idriss, and E. Ng, "Duration Characteristics 37 of Horizontal Components of Strong-Motion Earthquake Records, " 2.5.2-20

NUMAACQ>mments March 18, 1993 Une ln/Une Out 1 Bulletin of the Seismological Society America, Vol. 68, pp. 2 1487-1520, 1978. 3 49. B. A. Bolt, "Duration of Strong Ground Motion, 11 Proceedings of 4 the Fifth world conference on Earthquake Engineering, 1973. 5 so. w. w. Hays, "Procedures for Estimating Earthquake Ground 6 Motions," Professional Paper 1114, U.S. Geological Survey, 7 1980. -1 8 51. H. Bolton Seed et al. , "Representation of Irregular Stress 9 Time Histories by Equivalent Uniform Stress Series in 10 Liquefaction Analysis," National Science Foundation, Report EERC 75-29, October 1975. 12 52. s. T. Algermissen et al., "Probabilistic Estimate of Maximum 13 Acceleration and Velocity in Rock in the Contiguous United 14 States," U. s. Geological Survey Open-File Report 82-1033, 15 1982. 16 53. OSNRC, "Safety Evaluation Report Related to the Operation of 17 Diablo canyon Nuclear Power Plant, Units 1 and 2, 11 NOREG-0675, 18 supplement No. 34, June 1991. 19 54. OSNRC, "Identification and Characterization of Seismic

~~     ~e:r:::~=:~::"~:.w*a 2.s.2-21

NUMARC Comments March 19, 1993 Une ln/Une Out 1 APPENDIX F 2 3 PROCEDURE TO DETERMINE THE SAFE SHUTDOWN EARTHQUAKE GROUND RESPONSE SPECTRUM 4 5 6 F.l Introduction 7 8 This appendix describes an acceptable procedure to determine the safe 9 shutdown earthquake (SSE) ground response spectrum. The ground response -~ 12 13 spectrum is defined in terms of the horizontal and vertical motion at the free-field ground surface at the plant site. It is developed with the consideration of site seismic wave transmission effects and the mean magnitude and distance of earthquakes that produce the SSE (see Appendix C of this 14 Regulatory Guide). 15 16 The SSE for a site is determined by the procedure described in Appendix 17 B to this Regulatory Guide for the average spectral acceleration between 5-10 18 Hz and 1-2.5 Hz. Two frequency ranges are considered to insure that a wide 19 range of frequencies are considered in the development of the SSE response 20 spectrum. The procedure in Appendix B determines the SSE that is consistent -~ 23 24 with the design level for existing plants as of [Effective Date of the Final Rule]. The SSE ground response spectrum is determined by scaling a spectral shape to the levels for 5-10 Hz and 1-2.5 Hz. It is anticipated that a future Regulatory Guide will provide guidance in the assessment of site-specific 25 response spectra. 26 27 F.2 Procedure 28 29 The SSE ground response spectrum is determined by scaling an 84 th 30 percentile response spectrum shape to the average spectral acceleration levels 31 corresponding to the Reference Probability as defined in Appendix B to this 32 Regulatory Guide. The steps in the procedure are: F-1

NUMARC Comments March 19, 1993 Une ln/Une Out l Step I. Determine the average spectral acceleration level for 5-10 Hz and 2 1-2.5 Hz from the median hazard curves at the Reference 3 Probability (see Appendix B). 4 5 Step 2. Determine an 84 th percentile response spectrum shape for the site. 6 A site-specific or standard response spectrum shape can be used 7 (Ref. IF). Two response spectrum shapes are required; one that is 8 scaled to the 5-10 Hz (7.5 Hz) average spectral acceleration and -~ 11 12 another which is scaled to the 1-2.5 Hz level (1.75 Hz). Subsection F.3 identifies acceptable procedures to determine the response spectrum shape. 13 A standard response spectrum refers to a ground response spectrum 14 that is independent of earthquake magnitude and distance and that 15 accounts for site conditions in terms of general site categories 16 (e.g., rock or soil}. 17 18 Step 3. The response spectrum shapes determined in Step 2 are scaled to 19 the 5-10 Hz and 1-2.5 Hz average SSE levels, respectively from -~ 22 23 Step 4. Step I. This step is illustrated in Figure F.l. Step 3 produces two response spectra, one scaled to the average spectral acceleration between 5-10 Hz and the other between 1-2.5 24 Hz. For purposes of the site characterization, the applicant can 25 envelope the two spectra or alternatively, elect to retain two 26 spectra that are considered in design evaluations. The later 27 approach may be preferred when the response spectra have different 28 spectral shapes. This is illustrated in Figure F.2. 29 30 F.3 Ground Response Spectrum Shape 31 32 The response spectrum shape for a site should be developed considering F-2

NUMARC Comments March 19, 1993 Line In/Line Out I the response of surficial soil deposits to earthquake ground motion and the 2 mean magnitude and distance of earthquakes that produce the SSE. 3 Alternatively, a standard response spectrum shape of the type used in past 4 nuclear power plant designs is acceptable (Ref. IF). Currently, the Electric 5 Power Research Institute (EPRI) is completing work on the development of a 6 ground motion model for sites (rock and soil) in the stable continental region 7 (SCR) (Ref. 7F). It is anticipated that this work will be incorporated in a 8 future Regulatory Guide that provides specific guidance in the assessment of -~ 11 12 site-specific ground motion in the SCR. Site-Specific Response Spectrum 13 The development of a site-specific response spectrum shape should 14 consider the site soil and foundation properties, regional seismic wave 15 propagation and the mean magnitude and distance of earthquakes for the SSE. A 16 response spectrum .shape is determined for the mean magnitude and distance for 17 each frequency (i.e., 5-10 Hz and 1-2.5 Hz). The procedure to determine the 18 mean magnitude and distance is described in Appendix C of this Regulatory 19 Guide. -~ 22 23 Methods to determine a site-specific response spectrum include: Development of a database of strong motion records that are 24 selected to have magnitude and distance characteristics similar to 25 the mean magnitude and distance for the SSE. In addition the 26 strong motion records should have similar earthquake source 27 characteristics, propagation path, and recording site properties. 28 From the database of strong motion records an empirical response 29 spectrum shape can be developed (Ref. 2F). The 84 th percentile 30 response spectrum shape is used in the assessment of the SSE 31 ground response spectrum. While this approach may be preferred 32 for some sites, it does not explicitly account for randomness and F-3

NUMARC Comments March 19, 1993 Une ln/Une Out 1 uncertainty in ground motion. 2 3 2. A response spectrum shape can be developed using theoretical-4 empirical modeling techniques. These methods can be used to model 5 conditions that are not well represented in the strong motion 6 record database and to fully model randomness and uncertainty . 7 The EPRI ground motion moael for the SCR is such an example (Ref. 8 7F). The 84 th percentile response spectrum shape may be used in

• ~

11 12 3. the assessment of the SSE ground response spectrum . Analytical methods can be used to model the response of local site soil conditions (Ref . 3F-6F) to earthquake ground motions . The 13 response of the site soils should be evaluated for earthquake 14 motions defined at free-field ground surface on rock associated 15 with the mean magnitude and distance determined for the two 16 frequency ranges considered (e .g. , 5-10 Hz and 1-2.5 Hz). 17 18 Standard Response Spectrum Shape 19 It is acceptable to use a standard response spectrum shape to determine ~~ the SSE. However, since existing shapes were normalized to the peak ground 22 acceleration, they should be regenerated based on a scaling to the average 23 spectral acceleration for 5-10 Hz and 1-2.5 Hz. F-4

NUMARC Comments March 19, 1993 Une ln/Une Out 1 REFERENCES 2 3 IF. U.S. Nuclear Regulatory Conwnission, Design Response Spectra For Seismic 4 Design For Nuclear Power Plants," Regulatory Guide 1.60. 5 6 2F. U.S. Nuclear Regulatory Commission, "Development of Site-Specific 7 Response Spectra," NUREG/CR-4861, 1980. 8 -~ 11 12 3F. P. B. Schnabel, J. Lysmer, and H. B. Seed, "SHAKE-A Computer Program for Earthquake Response Analysis of Horizontally Layered Sites," Report No. EERC 72-12, Earthquake Engineering Research Center, University of California, Berkeley, 1972. 13 14 4F. I. V. Constantopoulos, "Amplification Studies for a Nonlinear Hysteretic 15 Soil Model," Report No. R73-46, Department of Civil Engineering, 16 Massachusetts Institute of Technology, 1973 . 17 18 SF. V. L. Streeter, E. B. Wylie, and F. E. Richart, "Soil Motion Computation 19 by Characteristics Methods," Proceedings American Society of Civil -~ 22 23 6F. Engineers, Journal of the Geotechnical Engineering Division, Vol. 100, pp. 247-263, 1974. W. B. Joyner and A. T. F. Chen, "Calculations of Nonlinear Ground 24 Response in Earthquakes," Bulletin Seismological Society of America, 25 Vol. 65, pp. 1315-1336, 1975. 26 27 7F. Electric Power Research Institute, "Guidelines for Determining Des ign 28 Basis Ground Motions," EPRI Report TR-102293, Vols. 1-4, May 1993. F-5

NUMARC Comments March 19, 1993 Une ln/Une Out 1 2 3 4 C Average Spectral Accelerations 5 Corresponding to the Reference Probability 6 7 8 - - Sa {RP)s-10

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• Frequency (Hz) 22 23 24 Figure F. l Il lustration of the procedure to scale a response spectrum shape 25 to the 5-10 Hz and 1-2 . 5 Hz average spectral acceleration levels corresponding 26 to the Reference Probability.

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NUMARC Comments March 19, 1993 Une ln/Une Out 1 Option 1 - Envelope C C 0

    ~

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( .) 1/ Q Q) Q) 0.. 0.. Cl) Cl) 1.75 7.5 7.5 Frequency (Hz) Frequency (Hz) Option 2 - Retain 2 Spectra C: 0 C'O

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( .) Q) 0.. C/) 1.75 7.5 2 Frequency (Hz) 3 4 5 6 Figure F.2 Illustration of the options to determine the SSE response 7 spectrum. F-7

DOCKET NlMBER" 50 rRoPoseo RULE l {_ 51 f (2.. J./ 11[0?) Direzione Centrale Sicurezza Nucleare e Protezione Sanitaria II Direttore Prat. n. 3160 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

Subject:

Proposed rulemaking 10CFR Parts 50, 52 and 100, "Reactor Siting Criteria" (Federal Register Vol. 57, No 203 - October 20, 1992).

Dear Sirs,

Referring to the proposed rules in the subject, I'm pleased to provide you with the attached ENEA-DISP comments. Sincerely yours a *\9~~~ Giovanni N aschi Encl. 1 Acknowledged by card .. Ml\Y...ll.8...:: - ENEA Comitato Nazi per la ricer per lo sviluppo Via Vitaliano Brancati 48 Fae-Simile: (06) 5013.429 dell'Energi ucleare mA 00144 Roma Telegrafo: ENEUR

• ROMA e delle En rgie Alternative Ento per la Nucv,;i Tocnologte. Telefono (06) 5007.1 Telex 612167 ENEUR I t'Ener£i3 o I' 1'rr:bi.:r1!D

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TH G'-~~,_ Off ICE OF . E SECRETARV OF THE COMMISSION Document SlallslfCS Postmark Dall :, /1-6 /q3 rd--)( J D"'- :J I/)_ '-I /OJ} Copies Received. _ _~ - : - - - - - - Add'I Copies Reproduced H Sp~ ;a: Distribution (l_J;tJ _'.5..__£__,.... iflL,

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Direzione Centrale Sicurezza Nucleare e Protezione Sanitaria Comments to: REACTOR SITE CRITERIA 10 CFR Parts 50. 52 and 100 (Proposed Rules in Federal Register October 20. 1992) US Regulations, especially at the CFR level, have always played a relevant role in public debates that took place in our country; this is the main reason why we have to be very reactive to those parts of the proposed rules which could have a heavy impact on this debate. In this respect the population density criterium (part 100.21) appears to be of major concern. In fact, if the figures reported into the rule, such as 500 people per square mile, are understood as Technical Reference, the future for nuclear energy would be even more difficult in Italy, as the averaged population density in our country is very close to the quoted figure. We realize that such limit is written down in a form (should) which is not mandatory, however we are afraid that a misunderstanding is still e possible. ~ On the other side, the Commission itself states that "nuclear power a: (.) plants meeting current safety standards could be safely located at sites significantly more dense than 500 people per square mile", that reflects our position. In conclusion, our suggestion is to remove the specification of numerical limits. Furthermore we think that such limits could find a better place in Regulatory Guides, where possible application criteria could be addressed, avoiding any possible misunderstanding and any arbitrariness in interpretation. We agree that remote siting is part of the concept of defence in depth and that sites with as low as possible population densities have to be preferred; however we believe that very strict requirements could be too heavy if compared with expected accidental behaviour and could discourage plant improvements. In this respect we suggest to adopt more realistic criteria such as distance based .weighting factors and limits applicable to sectors. Regarding the other questions, we believe that most of them arise from the policy of separating siting from designing. This is clearly the case of the problems connected to the definition of the exclusion area. On this matter our opinion is that both Exclusion Area and Emergency Planning should be correlated to reactor design and related safety feature. Other questions, such as the consideration of meteorological aspects ENEA Via Vitaliano Brancati, 48 Fae-Simile: (06) 30486393 Ente per le Nuove tecnologie, 00144 Roma Telegrafo: ENEUR ROMA !'Energia e l'Ambiente Telefono (06) 50071 Telex 612167 ENEUR I

Foglio n . .......... ?.............................. .. at site permit stage, depend on the specific characteristics of the available sites. The collection and the analysis of meteorological data are required at that stage, in our country, in order to exclude very unfavorable sites. Also from geology and seismology point of view, the separation between the two phases: siting and design, is not always simple. Particular analyses, for example the bearing capacity, cannot be treated separately. Such analyses, sometime, can exclude the site both from the economic point of view and for the feasibility of particular earthworks (es. deep backfitting in areas where shallow confined acquifers are present). Regarding the SSE, in the text it is alvays stated to use in a combined manner the probabilistic (first) and deterministic (second) approaches. This approach may be good in US because of the existance of EUS and WUS tectonic situations. However the majority of the countries in the world have only or interplate or intraplate situation. It would be better to state clearly that probabilistic methods are more important for in trap late areas while the deterministic methods for the interplate ones. Generally speaking the probabilistic methods provide too many data that can be misused. For the seismic input, 0, 1 g and a wide band spectrum are considered good reference ( i.e. minimum value) for SSE. Regarding capable and active faults, the new definitions appear very good. However the new terminology, respectively: capable tectonic source and seismogenic source, is not necessary absolutely. It is still not clear to many international Experts the difference between the old terms; therefore new terms can create only a larger confusion. The emphasis that is given to the importance of the "professional judgment" in taking decisions appears good. However it is important to put emphasis also on the fact that the judgment can be assessed properly when good Experts and good databases are available.

Wll.UAM L. Sl'EwAKT FT,.. U BEA

                                           ,..                              Innsbrook Technical Center Senior Vice President f                                            5000 Dominion Boulevard Glen Allen, Virginia 23060 804-273-3551
                                                *93 t1   25 P3 :08 I VI VIRGINIA POWER March 22, 1993 Secretary of the Commission                                          Serial No.: 92-723 U. S. Nuclear Regulatory Commission                                 NL&P/RBP R1 Washington, D.C. 20555 Attention: Docketing and Service Branch Gentlemen:

COMMENTS ON PROPOSED RULE CHANGES TO 10 CFR PARTS 50, 52 AND 100 REACTOR SITE CRITERIA Virginia Power has reviewed the Federal Register notice dated October 20, 1992 concerning reactor site criteria. The proposed rule, which is to apply only to future power plants, would provide new regulations regarding power reactor siting criteria, including maximum population density, minimum exclusion area distance and geologic, seismic, and earthquake engineering considerations. The purpose of this letter is to endorse the Nuclear Management and Resources Council (NUMARC) comments sent separately to the NRC and to provide specific comments regarding the proposed rule. - First, we agree with NUMARC that the proposed rulemaking may adversely affect public perception regarding the acceptable safety of existing plant sites during their operating term and especially during plant license renewal proceedings as a result of this proposed regulation. The duplicity of separate regulatory approaches for present and future plants is a contradiction of regulatory approaches. This proposed rulemaking will likely form the basis of future intervention and litigation for license renewal and future operating licenses due to the duality in regulation. In addition, since U.S. NRC regulation is frequently adapted by other nations, these issues could arise *internationally. Virginia Power strongly supports the NUMARC comments concerning the non-seismic portion of this rulemaking. We agree with the NUMARC conclusions and strongly recommend that radiological dose consequence evaluation factors contained in the current 1o CFR Part 100 be retained as the key determinant of site suitability. The proposed approach to establish projected population density as a criteria for use in assessing the suitability of future nuclear plants is not a precise science lending itself to a definitive conclusion. This approach is likely to be subject to protracted "expert" contention and litigation which may effectively preclude siting or license renewal on a

wholly subjective basis. The present approach appropriately addresses exclusion distance and population density based on an appropriate technical basis which provides protection of public health and safety with an adequate defense in depth on a risk/safety basis. The proposed rule simply does not provide a technical basis for regulation. Finally, we agree with the NUMARC com ments on the seismic portion of this rulemaking which characterizes the requirem ent to conduct both a deterministic and probabilistic seismic evaluation as fundamentally flawed, since there are no clear technical means to reconcile differences between such evaluations. As mentioned above, this proposed approach also appears likewise destabilizing to the siting process. Should you have any questions, please contact us. Very truly yours, J)t~ W. L. Stewart cc: Mr. Ron Simard Director of Industry Relations and Admi nistration Division Nuclear Management and Resources Council 1776 Eye Street, N. W. Suite 300 Washington , D. C. 20006-2496

DOCKET NtJMBER" Southern Nuclear Operating Company Post Office Box 1295 Birmingham, Alabama 35201 P1 5 ~ -,1111, ;J--} tJ (J PROPOSED RULE.:..;......__..,....L-- Telephone 205 868-5086 C!i 7 F R. L/ I ro "';J..) -

                                                                               ,-i-tll1t J. D. Woodard Southern Nu~a @.~ating:_[ompany Vice President                                                                 the southern electric system Farley Project Mar ch 24, 1993 Docket Nos. 50-348 50-364 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Docketing and Service Branch Comments on Proposed Rule "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants" (57 Federal Register 47802 of October 20, 1992)

Dear Mr. Chilk:

Southern Nuclear Operating Company has reviewed the proposed rule, "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," published in the Federal Register on October 20, 1992. In accordance with the request for comments, Southern Nuclear Operating Company is in total agreement with the NUMARC comments which are to be provided to the NRC. In addition, we understand that certain foreign utilities have concerns that are consistent with NUMARC ' s positions, e.g., the lack of sound technical bases for the exclusion zone distance and the population density criteria. Should you have any questions, please advise. Respectfully submitted, JDW/JDK MAY t t 1993 --- Acknowledged by card .............- .......- ..

U.S. NUCLEAR REGULATORY COMMISSIC>ft DOCKETING & SERVICE SECTION Off ICE OF= 'THE SECFIETARV OF THE CQ:-At-AISSION 0,Jcument Smtistics

U. S. Regu l atory CoD111ission Page 2 cc: Southern Nuclear Operating Company R. D. Hill, Plant Manager U.S. Nuclear Regul atory Conwnission, Washington, D. C. G. F. Wunder, Licensing Project Manager, NRR U.S. Nuclear Regul atory Commission, Region II S. D. Ebneter, Regional Administrator G. F. Maxwell, Senior Resident Inspector State of Alabama Carole Samuelson, Acting State Public Health Officer

Georgia Power Company 40 Inverness Center Parkway Post Office Box 1295 DOCKET NUMBER PROPOSED RULE PR 5 4 51.. I J-l o O Birmingham, Alabama 35201 Telephone 205 sn-7279 (5" 7 F ~ L/ 7 ~(J 2-)  ;":, * -- L ~,j~t1/J' J. T. Beckham, Jr. *93 ** ,., Georgia Power Vice President - Nuclear Hatch Project the southern electnc system March 24, 1993 Docket Nos. 50- 321 50-424 HL-3212 50-366 50-425 ELV-05302 Mr. Samuel J. Chilk Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Docketing and Service Branch Comments on Proposed Rule "Reactor Si te Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants" (57 Federal Register 47802 of October 20, 1992)

Dear Mr. Chilk:

Georgia Power Company has reviewed the proposed rule, "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," published in the Federal Register on October 20, 1992. In accordance with the request for comments, Georgia Power Company is in total agreement with the NUMARC comments which are to be provided to the NRC. In addition, we understand that certain foreign utilities have concerns that are cons i stent with NUMARC's pos i tions , e.g., the lack of sound technical bases for the exclus i on zone distance and the population density criteria.

• Should you have any questions, please advise.

Respectfully submitted,

                                             ~     - T. Beckham,      r.

JTB/ JDK llfAY 11 f993 Acknowledged by card ....._._ ..................:::

U.S. NUCLE*.-1 RcGULATOAY COMMISSIOr-. DO*:KETING & SERVICE SECTION Ofr!CE OF THE SECRETARV OF THE COMMISSION Dccumont Statistics Postmari( Dall 3 / , __,., / q 3 F~ Cfr-- J /-z.'-l I~ 3 Ccc!iS Received I

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Georgia Power , \ U. S. Regulatory Commission Page 2 cc: Georgia Power Company C. K. McCoy, Vice President, Plant Vogtle W. B. Shipman, General Manager - Plant Vogtle H. L. Sumner, Jr., General Manager - Plant Hatch NORMS U. S. Nuclear Regulatory Commissi on, Washington, DC K. N. Jabbour, Licensing Project Manager - Hatch D. S. Hood, Licensing Project Manager - Vogtle U.S. Nuclear Regulatory Commission. Region II S. D. Ebneter, Regional Administrator L. D. Wert, Senior Resident Inspect or - Hatch B. R. Bonser, Senior Resident Inspector - Vogtle State of Georgia J. D. Tanner, Commissioner, Department of Natural Resources HL-3212 ELV-05302 700775

STATE OF CALIFORNIA -THE RESOURCES AGENCY PETE WILSON, Governor DEPARTMENT OF CONSERVA ~Q~"FT 11 ._, . t) t) DIVISION OF MINES AND GEO QGY- . ~ S f9 --5 1-_ r} *

~~~s~~~tMEt1~~ARTERS                   ( 5" 7 Fr<.. 41 h, 1.               , - ,_:;

Sacramento, CA 95814-3531 Phone (916) 445-1923 Fax (916) 445-5718

                                                                   *93 MA , 24 P3 :53 BBS     (916) 327-1208 l,

Mr. Andrew J . Murphy, Chief Structural and Seismic Engineering Branch Division of Engineering Office of Nuclear Regulatory Research Dear Sir; I commend the Nuclear Regulatory Commission for undertaking the revision of the existing criteria for the seismic and geologic siting of nuclear power plants (Appendix A to Part 100 CFR 10). The proposed Appendix B makes some substantial improvements. The following comments are provided by me in my capacity as Chair of the Geologic Hazards Committee of the Association of American State Geologists. The State Geologists of several other states are making coordinated responses. The material presented in this letter, however, does not constitute recommendations from the entire Association membership because I have not had sufficient time to circulate this item for concurrence. 0 Taken together with Draft Regul atory Guide DG-1015, the Appendix B guidance for evaluation appears adequate for the deterministic safe shutdown earthquake ground motion. However, the controversies aris ing from disagreements over the delineation of seismotectonic provinces that developed among parties under the provisions of Appendix A, will not be significantly reduced by issuance of Appendix B because the definition of " seismogenic source" is as vague and inclusive as "seismotectonic province. 11 Enough has been learned in the 20 years since Appendix A was developed to establish some test of "uniform earthquake potential. " Providing some criteria as guidance, but leaving the possibilities open to reasonable a dd itions woul d increase the ef_f iciency of further siting deliberations. Associations of geology and seismicity should be stressed, such as demonstrated by blind thrust structures or areas with *structural styles and b values that contrast with their surrounding regions. The correlation of seismici ty and geology rightfully received less emphasis 20 years ago because there were fewer recording networks. Reflection data and deep wel l records were also less available. Provinces with both distinctive seismicity and geologic style are present, should be given greater weight than interpretations of seismogenic sources where only geology or seismicity alone do not characterize the areas as

                                                                                  !: ~. . . ,;

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0 Given the contrasts in availab l e data on seismicity and earthquake history in various portions of the United States, the use of both deterministic and probabilistic methods in analyses in conjunction is justified. Further guidance is needed, however, in using the results of these comparisons in the identification of the safe shutdown earthquake ground motion, site suitability, and seismic design requirements. 0 Determining the median for the annual probability of ground motion exceedence of the existing operating nuclear power plants in the us may help to assess general exposure. However, it is not clear that this value should be used in future siting requirements. As staff points out in DG-1015, an interim procedure is necessary to calculate the probabilistic seismic hazard east of the Rocky Mountains because the two current state-of-the-art methods do not give similar results. For the Western US, as also pointed out in DG-1015, a probabilistic data base is not available, so that no procedure exists to compute the median SSE motion for these sites. Even if it were possible to calculate the median SSE value for existing sites, it does not necessary follow that th i s va l ue should constitute a standard for which new plants having lower probability estimates should be presumed to b e acceptable. The logic of such a standard is not clear. Does this meant that it has been established that all existing nuclear plants have been engineered to have the capacity to be safely shutdown following their design earthquakes? Such a conclusion has not been empirically demonstrated by plant performances during earthquakes. Furthermore, since existing nuclear plants are not distributed in such a way as to comprise a weighted representation of seismic potential in the us, the median annual probability of ground motion exceedence at these sites does not constitute a proven standard or reference by which one can establish an upper-bound for design requirements or for acceptable site conditions. 0 The minimum value due of O.lg horizontal peak ground acceleration referred to at Appendix B V (c) is appropriate. There is no place in the United States where the occurrence of magnitude (M) 5 size earthquakes can be safely ru led out. Furthermore, in some localities it is not poss ible to identify the sources that could be associated with future earthquakes. Since peak ground acceleration on rock will exceed O.lg in the Western US at distances closer than 13km to a MS sources (Joyner and Bohr) and this level of shaking extends out to even greater distances east of the Rocky Mountains, it seems reasonable to consider ground motion O.lg a lower bound. A more conservative, bu t also defensible,

Mr. Andrew J. Murphy Page 3 March 23, 1993 approach would be to use o. 2 g ( approximately the maximum ground motion at the surface above a M5 source) as the lower bound. O The largest historical earthquake should not be considered to be the Deterministic Source Earthquake except where there are no geolog ic circumstances that can reasonably be interpreted to be sources of larger events. In such cases, more weight should be given to probabilistic analysis method it if results in a larger SSE ground motion than the value derived by the deterministic method. O The II footprint II of a nuclear plant should never be placed at a location on the mapped trace of capable tectonic source that can experience surface faulting. Thank you for the opportunity to review the draft regulations. Sincerely, James F. Davis State Geo logist cc: Members, Geologic Hazards AASG

Telephone (508) 779-6711 YANKEE ATOMIC ELECTRIC COMPANY TWX 710-380-7619 ~------ \__)(J ~ *"\ 580 Main Street, Bolton, Massachusetts 0 1740-1 ;;8 , ,

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                                                                        *93 HAR 24 P2 :53 PROPOSED Rl!L2      t D  O
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(S- 7 FR k/ 7</;'01-) March 23, 1993 FYC 93-004 SPS 93-027 Secretary of the Commission U. s. Nuclear Regulatory Commission Washington, DC 20555 Attention: Docketing and Service Branch

Subject:

Request for Comments Concerning Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants (57FR66457)

Dear Sir:

Yankee Atomic Electric Company {YAEC) appreciates the opportunity to comment on the subject Federal Register Notice. Yankee is the owner of the nuclear power plant in Rowe, Massachusetts and our Nuclear Services Di vision also provides engineering and licensing services to other nuclear power plants in the United States. These comments are also filed on behalf of the Vermont Yankee Nuclear Power Corporation and the Maine Yankee Atomic Power Company. The detailed comments concerning the seismic siting criteria were developed in a collaborative effort by YAEC, Connecticut Yankee Atomic Power Company {CYAPCO) and Northeast Nuclear Energy Company (NNECO). Yankee is providing responses to the specific NRC questions concerning nonseismic reactor siting criteria in Attachment I to this letter. Attachment II provides comments on the Proposed Regulatory Guides DG-1015, DG-1016, and DG-1017. In addition to requesting comments on the proposed Regulatory Guides, the staff has raised several questions regardi ng its proposed process for determining the SSE. Detailed responses to questions on DG-1 015, along with an alternative process for determining the SSE are included in Attachment II. We consider any attempt to provide comments on SRP-2.5.2 at this time as pointless; when a credible, workable process to determine the SSE is offered, a significant rewrite of SRP-2. 5. 2 would be expected. With respect to the proposed Appendix B to 10CFRl00, we fully endorse the comments filed by the Nuclear Management and Resources Council (NUMARC). Acknowledged by card .~M<<r..J..!.~.....::

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Secretary of the Commission March 23, 1993 Page 2 As a general comment on the staff's siting proposals, there is neither a discernable technical basis nor a credible nexus to the commission's Safety Goals in these proposals. Introducing additional "safety" conservatisms which have no demonstrable technical basis into the siting process will likely have the effect of ruling out of consideration sites that might well he superior from the standpoint of general environmental (non-accident) concerns and yet completely appropriate form the standpoint of safety concerns. Furthermore, these proposals inappropriately call into question the safety and acceptability of existing plant sites. We hope that the Commission and the staff will take full advantage of the opportunity offered by this rulemaking and in the next draft of this proposal show that it is possible to establish a meaningful linkage between the Commission's rules and its Safety Goals. sincerely yours, rf)d D. W. Edwards Director, Industry Affairs /sf Attachments

ATrACBMENTI Response to Questions Related to Nonseismic Reactor Siting Criteria

1. Should the Commission grandfather existing reactor sites having an exclusion area distance 0.4 miles (640 meters) for the possible placement of additional units. if those sites are found suitable from safety consideration?

Answer: Currently operating pl ant sites have demonstrated acceptable safety for current reactor designs. Once approved the site should never be challenged based upon later interpretation of mi nor aspects of the rule. The placement of additional units of advanced design on these sites should be determined on the basis that safety is maintained as a result of operating all the licensed units on a site. Expected dose is the measure that has been used very effectively to date. That same basis should be utilized for determining acceptability of unit placement on a site not occupied by an existing unit.

2. Should exclusion area distance be smaller than 0.4 mile (640 meters) per plant having reactor power levels significantly less than 3800 megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example. 0.25 miles or 400 meters)?

Answer: Distance is an inappropriate parameter to use for siting. Exclusionary distance should be determined based upon radiological dose consequence evaluation as is required by the present 10CFR Part 100. Exclusionary distances less than the 0.4 miles proposed. have been found by the NRC to be adequate for the protection of public health and safety for 25 of the 75 currently operating sites. Flexibility should be retained for future sites to establish exclusion area size smaller than the 0.4 miles when a small area provides adequate protection. 3A. Should numerical values of population density appear in the regulation or should the regulation provide merely general guidance with numerical values provided in a Regulatory Guide? Answer: In this case. neither option offered is acceptable. Appropriate determinants for site acceptabil~ty should remain the radiological dose consequence evaluation factors as are specifically presented in the current 10CFR Part 100. These arbitrary numbers intended to discourage metropolitan siting of nuclear power plants have no credible technical basis or demonstrated linkage to the NRC*s Safety Goals. C76\806

3B. Assuming numerical values are to be codified, are the values of 500 person per square mile at the time of site approval and 1,000 persons per square mile 40 years thereafter appropriate? If not, what other numerical values should be codified as the basis for these values? Answer: Since population density limits are not the key detriments of off-site radiological dose risk, requiring the reconsideration of population density criteria during early site permit renewal is not appropriate. As stated previously, these values should not be considered as the upper limit of acceptability. 3C. Should population density criteria be specified out to a distance other than 30 miles (50 km), for example, 20 miles (32 km)? If a different distance is recommended, what is its basis? Answer: Answer as stated above, there is no justification for any particular distance for setting population density criteria.

4. Should the Commission approve sites that exceed the proposed population values of 10CFRl00.21, and if so, under what condition?

Answer: The Commission should approve sites that the applicant has successfully demonstrated adequately protect health and safety using the current radi ol ogi cal dose consequence evaluation factor. There is no other technically defensible ulterior other than risk.

5. Should holders of early site permits, construction permits, and operating license permits, be required to periodically report changes in potential off-site hazards (for example, every five years within five miles)? If so, what Regulatory purpose would such reporting requirements serve?

Answer: The reporting requirement is redundant to current requirements for operating licenses to report potential off-site hazard impacts on the plant to the degree the impact affects public health and safety. 10CFR50.21Ce) requires an annual update and report to the NRC for the final safety analysis report (FSAR). There seems to be little if any Regulatory benefit for many of the reports required of a licensee.

6. What continuing Regulatory significance should the safety requirements in 10CFR Part 100 have after grantiig initial operating license or combined operating license under 10CFR Part 52?

Answer: 10CFR Part 100 safety requirements should not have any continuing Regulatory significance beyond the issuance of site permits or the siting portion of combined operating licenses. C76\805

7. Are there certain site metallurgical conditions that should preclude the siting of a nuclear power plant? If so. what are the conditions that cannot be adequately compensated for by design features?

Answer: We are not aware of any metallurgical conditions that cannot be adequately compensated for by design features.

8. In the description of disposition of the recommendations of the site policy task force report {NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to consider any recommendation not adopted and. if so, what are the basis for reconsideration?

Answer: We are unaware of any obligation for the Commission to accept any, let alone all. the recommendations of a task force that it constitutes study and issue. In the particular case of the siting policy task force we are unable to identify any additional recommendations that should be reconsidered for adoption. Furthermore, it is unfortunate that a clearly outdated and technically questionable report is given as much credence as has been the case in the staff's dealing with siting issues. C78\30II

Attachment II Comments on 10CFR100 Appendix B, Draft Regulatory Guide DG-1015 ouestion Nwnber 1 In making use of both *deterministic and probabilistic evaluations, how should they be combined or weighted; that is, should one domi-nate the other? (The NRC staff feels strongly that deterministic investigations and their use in the development and evaluation of the Safe Shutdown Earthquake Ground Motion should remain an impor-tant aspect of the siting regulations for nuclear power plants for the foreseeable future. The NRC staff also feels that probabilis-tic seismic hazard assessment methodologies have reached a level of maturity to warrant a specific role in siting regulations.) Answer to oues~ion 1 1.0 USNRC Draft Siting Approach (DG-1015)/Background The draft 10CFRl00 Appendix B siting approach, as defined in DG-1015, retains the deterministic-requirement of Appendix A 10CFR Part 100 and adds a parallel probabilistic methodology requirement as the basis for assessing Safe Shutdown Earthquake (SSE) ground motions. The basis for adopting this approach can be traced to a report published by the National Research Council Committee on Seismology (1). A recommendation of this committee. i s :

      "At   low   probability    levels,   both   deterministic  and probabilistic hazard studies should be conducted to arrive at the appropriate   seismic    design or evaluation      criteria. The probabilistic seismic hazard analysis should be deaggregated to determine the source, magnitude, and distance of the earthquakes that are dominating the hazard.       The probabilistic result can provide a quantitative basis for assessing the reasonableness implied by the deterministic estimate."

It should be noted that a trial application of. this optimistic approach is not presented in the report. Figure 1 illustrates the proposed NRC siting process. As can be seen the initial step in the process is to obtain the region and site-specific geological, seismological, and geophysical data. Given the results of these studies a deterministic evaluation is conducted and seismic sources and controlling earthquakes are determined for the site. This pathway is similar to the Appendix A 10 CFR Part 100 process used to license existing NPPs and therefore retains all of its difficulties. The initial step along the probabilistic path is to conduct an Electric Power Research Institute (2) or a Lawrence Livermore National Laboratory (3) seismic hazard analysis (SHA) of the site. 1

This SHA is based on use of inputs developed from teams of nationally recognized experts and specific to each methodology. The results of the SHA are then used to determine ground motions consistent with acceptable probabilities within a methodology. These acceptable probabilities are defined in DG-1015 Appendix B. They are: LLNL - 1. 0*10 4 EPRI - 3. 0*10-5

• The hazard, at that ground motion, is then deaggregated producing results in terms of a mean magnitude and distance. Mean magnitudes and distances are calculated for the peak ground acceleration, and at two frequencies ( about 7 hz and 2 hz) . Based upon these results, controlling earthquakes are developed.

The final step in the NRC approach is to compare the controlling earthquakes developed in both the deterministic and probabilistic evaluations. The SSE is defined by the envelope of the controlling earthquakes and is considered adequate if it compares favorably to the above specified limits. r 1.1 significant Issues Associated with the Deterministic Pathway Relative to Future Licensing Historically, the SSE at existing EUS NPPs has been developed based on the tectonic province approach. The tectonic province approach was adopted to accommodate the lack of understanding of the earthquake generating mechanism and the lack of data in the EUS. Where earthquakes could not be related to a geologic structure (few can be clearly correlated with geologic ~tructures, evidence the Charleston issue) they were associated with tectonic provinces. The SSE at the site was determined by assuming that the largest historical earthquake in adjacent provinces occur at the boundary of the seismotectonic province at its closest point to the site. The largest historical earthquake within the province in which the site is located (host zone) is assumed to occur adjacent to the site. Numerous cases in the literature document the difficulties and inconsistencies associated with the tectonic province approach (4,5). For the northeast, Figure 2 shows the tectonic interpretations that were adopted by the staff in the licensing of Indian Point, Pilgrim, Millstone, and Seabrook (4). It is of interest to note that these plants were licensed well before the LLNL and EPRI studies were performed. Given the range of subjective interpretations concerning seismic sources for the EUS that is documented in both the LLNL and EPRI studies, it is not prudent to require a licensee to resolve an issue that is clearly not resolvable within the scientific community. In particular, this requirement should not be in a licensing document. Perusal of the many seismic source interpretations for a given location presented in the LLNL and EPRI studies by nationally recognized experts, highlights the profound differences that exists between these experts. Figures 3 and 4 2

illustrate the diversity of opinion concerning seismic source zones at two existing NPP sites. Only the host zones are shown on these figures. As can be seen, the NRC tectonic interpretations presented in Figure 2 are put into perspective in Figure 3. Another issue to be resolved is the determination of the deterministic source earthquake (DSE) for each seismic source zone. How do the definitions of the DSE used in the deterministic pathway and the upper bound magnitude estimates used in the probabilistic pathway differ? The DSE, as defined in DG-1015 (9), 'is the largest earthquake that can reasonably be expected to occur in a given seismic source in the current tectonic regime.' An important parameter in seismic hazard analyses is the upper limit of the range of magnitude values (Mu}. In the LLNL study (3), Mu is defined as, 'the upper limit for the distribution of earthquake magnitude within a zone. given the current tectonic and seismic conditions.' Figures 5 and 6 illustrate the range of upper bound magnitude estimates that for the host zones shown in Figures 3 and 4 could be applied as DSEs in a deterministic analysis given the present definition of the DSE. As can be seen, this application could result in design values in excess of 0.3g for these sites. The potential for misuse/abuse of the upper bound magnitude estimates from the LLNL and EPRI studies will invariably result in licensing instability. 1.2 Evaluation of the Stability of the Deterministic Process - Past Licensing Decisions - Role in Future SSE Determinations The Electric Power Research Institute (EPRI) and Lawrence Livermore National Labs (LLNL) have mature seismic hazard methodologies and both have calculated the seismic hazard at EUS NPP sites. Because the LLNL and EPRI methodologies each use internally consistent- data for calculating seismic hazard at all sites, i.e., common attenuation models, experts, and calculational procedures, the relative seismic hazard at various NPP sites for each methodology is easily determined. In this context, the seismic hazard results from these studies can be used to independently evaluate the probability of exceeding the current licensing basis at existing NPPs. In particular, this comparison evaluates the consistency ( in terms of probability of exceedance) of the deterministic licensing process that has been used to define the current licensing basis at existing EUS NPP sites. A fundamental licensing/engineering premise is that the seismic design basis (defined in Appendix A as the Safe Shutdown Earthquake, SSE) should be proportional to the expected seismic loadings; in other words, the higher the expected seismic loadings, the higher the seismic design basis. Figure 7 presents the PGA values associated with the SSEs determined for 61 EUS NPP sites. As can be seen, these seismic design levels vary between o. lg and 0.25g. Clearly, if the deterministic process used to determine these seismic design levels is consistent with the above premise 3

(i.e. , higher seismic loadings require a higher seismic design level) then it should follow that the probability of exceeding each plant's seismic design level should be about the same. Figure a is a plot of the probability of exceeding the Safe Shutdown Earthquake (SSE) at 61 EUS NPP sites based upon median hazard results from both LLNL and EPRI. Analyses have shown (6) that the median results (as compared to mean, 15th or 85th percentile results) are most consistent between LLNL and EPRI. fil! oan be seen. the probability of exceeding the SSE from site to site is far from consistent within each methodology (LLNL or EPRI). These results demonstrate that the current deterministic approach does not define consistent seismic design levels (in terms of probability of exceedance) from site to site. However, what also can be seen from Figure 8 is that the trend from site to site between the LLNL and EPRI results is consistent in a relative sense, meaning that both studies consistently identify high and low hazard sites, albeit the magnitude of the absolute probability of exceeding the SSEs vary significantly. As an aside, this wide variation in probability of exceedance does not imply that those plants with the highest probability of exceeding their seismic design basis are unsafe. PRAs and seismic margins studies have shown that plant capacity exists well beyond the SSE and that the major contribution to seismically-induced core melt frequency for EUS plants comes from earthquakes that are about 2 to 4 times the SSE (7). Therefore, precise determination of the SSE is not critical to safety given that conservative criteria are used in the probabilistic pathway to determine the site SSE. Figures 7 and 8 highlight the inability of the deterministic licensing process to define consistent design levels from site to site. Because the LLNL and EPRI results define the seismic hazard at the PGA and frequencies of 25, 10, 5, 2.5, and 1 Hz, it is possible to convert SSE response spectra at the PGA and each frequency to probabilities of exceeding the plant seismic design basis. Figure 9 presents a plot of the probability of exceeding the plant design basis using the LLNL results for EUS NPP sites. It dramatically illustrates the inconsistency in probability of exceedance between sites not only at the PGA but also at various frequencies. This figure also, shows that for a given spectrum, the probability of exceeding the design basis typically varies by about two orders of magnitude between about 25 hz and 1 hz. Based on the above, we conclude that across the current population of NPPs the deterministic process used to define the seismic design basis at existing Nl?Ps bas not resulted in consistent seismic design levels (as defined by similar probabilities of exceedance) and should be eliminated as a licensing mechanism to determine the SSE for future plants. Also, because the probaJ:>ilistio approach gives consistency in terms of probability of exceedance between sites, and the deterministic progess does not, the dual pathway assumption of comparability is invalid. In particular, favorable comparisons between the deterministic and probabilistic estimates ot the SSB can only be considered as fortuitous, not confirmatory. 4

1.3 Role of the Propa];>ilistio Pathway in Future Licensing The process outlined in Figure 1 shows that the first step in the probabilistic pathway is to perform a LLNL or EPRI SHA. The next step is to combine the 5 and 10 hz median hazard curves and based upon the specified acceptable probabilities, determine the ground motion for use in deaggregation. This results in a mean magnitude and distance associated with the acceptable site ground motion. Mean magnitudes and distances are also determined for the Peak Ground Acceleration (PGA) and at the mean of the 1 and 2. 5 hz hazard curves using the acceptable probability associated with the 5 and 10 hz values. Given the acceptable probability consistent with the methodology, the acceptable ground motion at the site can be determined, deaggregated, and the mean magnitude and distance determined. controlling earthquakes based on probabilistic results are then developed and compared against the controlling earthquakes developed in the deterministic pathway. The SSE is defined by the envelop of the probabilistic and deterministic results. Interestingly, DG-1015 Appendix B states, 'The SSE is adequate when the probability of exceeding the SSE compares favorably to the limits shown in these figures.' Given this condition, it is not necessary to deaggregate, develop controlling earthquakes, compare with the deterministic controlling earthquake, and then finally define the SSE. Because there are large differences in the absolute value of the probabilities determined for a site between methodologies (see Figure 8), and because these probabilistic estimates are from valid studies, the NRC has adopted the approach of using the results of the LLNL and EPRI studies in a relative manner rather than as absolute probabilities. Adding to the basis for using these probabilities in a relative sense is the awareness that these probabilistic results are highly dependent on 1 expert opinion.' The strength of this approach is that once the methodology has been calibrated to a standard, traditional arguments about 'correct 1 inputs are in a relative sense unimportant, as long as the inputs are applied consistently from site to site. Also, because a data base of inputs from nationally recognized experts exists for the entire EUS for both methodologies, reproducibility of results for each methodology is insured. In effect, use of these standardized inputs results in an implie4 map of aoceptal:>le ground motion values. such a map would certainly enhance predictability and is consistent with past recommendations (8). Development of this map will simplify the licensing process and should be a priority for both the NRC and industry. The probabilistic pathway, as defined in DG-1015, is built around the strength associated with the use of the probabilistic results in a relative manner. Given the policy decision to use these results in a relative manner, only a basis for defining acceptable probabilities within a methodology need be defined. Presently, the NRC has concluded that existing NPP SSE spectra should be used to calibrate the methodologies. In particular, if the SSE at some future site is lower in probability than the median probability of exceeding current plant spectra, then the SSE is considered acceptable. The dashed line in Figure 9 defines the median 5

probability relative to existing sites for the LLNL results. We have evaluated this process at existing sites and has found that because the current NRC approach allows use of either the LLNL or EPRI methodologies, different results are produced. This is viewed as a weakness in terms of licensing. Given that the acceptable probabilities are to be based on the median probability of exceeding current spectra, it is recommended that both methodologies be used, and that the higher of the two be used as the acceptable ground motion value to determine the site SSE. Use of both LLNL and EPRI is a conservative 'sanity check' using comparable pathways. Figure 10 shows the differences in SSE at existing sites, given the NRC approach. The values shown in Figure 10 are based upon scaling a R.G. 1.60 spectrum to the acceptable ground motion, assuming that ground motion applies at 9 hz. The 9 hz value was chosen because it is a corner frequency when developing a R.G. 1.60 spectrum. As can be seen from Figure 10, use of the EPRI methodology can produce higher results than the LLNL methodology. An alternative method of calibrating the results from each methodology is available which essentially eliminates the above problem. We discuss this approach later in our comments. Summarizing, the lessons learned from the LLNL/EPRI studies:

1. There is no consensus concerning seismic sources for the EUS.

2. The deterministic licensing process has not resulted in consistent (in terms of probability of exceedance) design bases at existing EUS NPPs and should not be used to determine the site SSE for future plants.

3. There is internal consistency within the LLNL and EPRI methodologies for ranking the seismic hazard between sites.

4. There is relative consistency between these methodologies for ranking the seismic hazard at a site.

5. Median hazard.results give the most consistency between methodologies.

6. Within the hazard results for a site, there is a huge variation of results between experts

7. only probabilistic methods (LLNL and EPRI or a Resolution methodology) should be used in the determination of the SSE at future EUS sites.

6

1.4 Use of Site-specific Deterministic studies Figure 1 implies that results from the site-specific evaluations will be used as inputs to the LLNL or EPRI SHA. However, a reading of DG-1015 suggests that only documented inputs defined by the teams of experts are to be used in the probabilistic pathway. Scientifically, the process has always been to update analyses based upon 'new' information. However, acceptable probabilities defined in DG-1015 Appendix Bare based on the results of the LLNL and EPRI studies at existing NPPs. Given that the results of these studies are to be used in a relative manner, should the results of the deterministic evaluations be used to update the inputs to the probabilistic evaluation? A clear consequence of this decision to use the results of the LLNL and EPRI studies in a relative manner is that updating of the inputs is inconsistent with the philosophy of using the results in a relative sense - it is also destabilizing. Furthermore, attempts at updating will prove to be a prolific source of trouble in the licensing arena. In particular, it can be shown that for any given site, the range of seismic sources, seismici ty parameters and upper bound magnitude interpretations for each expert is large. Across all experts, these interpretations represent honest differences of opinion between experts. Figure 11 shows the typical range of results for a site by the experts in the LLNL study. Given that a utility decides to build a plant near this site, the choice of geotechniqal consultant may be expert c. Assuming the consultant defines new source zones and seismicity parameters based on site-specific information, how should this information be used in the* probabilistic analysis, or more appropriately, should it be used in the probabilistic analysis? Clearly, if inputs from consultant C

• are- assumed to be "truth", then application of the acceptable probabilities based upon all expe:i::ts will result in acceptable grounq motions that are low relative to what would be obtained if all the experts were used. The converse is true if expert B's inputs are assumed to be tr~e. Therefore, unless the inclusion of new data is based upon the interpretations of the original experts or teams - updating cannot be based only upon the interpretations of the utility's consultant. Allowing the evaluation of innumerable alternative input assumptions by the experts or teams will surely produce the same kind of paralysis and delay associated with the old Appendix A licensing process. Furthermore, there is no reason to not believe that given new information (paleoseismic for instance) about a region around some future site that similar honest differences of opinion will continue to be expressed by the experts.

In this context, there should be no attempt to update inputs to the LLNL or EPRI programs based upon site-specific evaluations. It should be remembered that the documented inputs to the LLNL and EPRI studies represent the best judgement of nationally recognized experts (LLNL) and Teams of Consultants (EPRI}. These methodologies in effect represent pseudo-tectonic provinces that 7

will produce an SSE estimate consistent with current NPP design values. Geotechnical findings at the site/site area should be evaluated deterministically just as was done for Appendix A plants. The probabilistic results should be viewed as if they reflect an Appendix A interpretation of tectonic provinces because, in the end, that is exactly what the probabilistic results are - a mechanism to handle all imagined source zones and come up with an answer that is reasonable. Therefore, updating of the inputs to LLNL or EPRI should not be performed on a site by site basis. Updating across all sites should be accomplished every 10 years. This recommendation is also consistent with DOE-STD-1024-92. Question Number 2 In making use of the probabilistic and deterministic evaluations as proposed in Draft Regulatory Guide DG-1015, is the proposed proce-dure in Appendix c to DG-1015 adequate to determine controlling earthquakes from the probabilistic analysis? Answer to Question 2 2.0 Determination of Controlling Earthquakes A specific issue for comment is the process defined in Appendix C of DG-1015 to determine controlling earthquakes. The process defined in DG-1015 requires the following:

1. Perform a LLNL or EPRI SHA

2. Based upon acceptable probabilities for that methodology determine acceptable ground motions at the PGA, about 7 hz and about 2 hz.

3. At these acceptable ground motion values, deaggregate the hazard and determine controlling earthquakes (mean magnitudes and distances) associated with these ground motions. *.

4. Compare these.controlling earthquakes (M,R pairs) with the

• results from the deterministic analysis.

5. Using standard Review Plan (SRP) section 2.5.2 determine the ground motions at the site given the controlling earthquakes.

6. The SSE is the envelope of the step 5 results.

7. The SSE is considered acceptable if the probability of exceeding the SSE at about 7 hz is less than the acceptable probabilities used in step 2.

The calculation of M,R pairs from the probabilistic analysis is required so that comparisons can be made between the probabilistic and deterministic pathways. SRP Section 2. 5. 2 is then used to determine ground motions at the site. As stated earlier, successful application of the Appendix A process is no longer possible, and therefore these comparisons (sanity check) are not necessary. Furthermore, there is no basis to believe that a favorable SSE comparison between the probabilistic and deterministic pathways is anything but fortuitous. Interestingly, comparisons presented to date show that the probabilistic mean magnitude and distance compares favorably with the magnitude and distance determined from Appendix A when the plant wos licensed {backward look). As stated above, in light of EPRI/LLNL studies, 8

 .successful application of the Appendix A process to a new plant (forward look) is not possible.

The mean magnitude and distance determined from the probabilistic process is not unique. For a given location it is totally dependent on the acceptable probability level. Figure 12 shows the relationship of mean magnitude and distance for the source zones shown on Figure 13 assuming the site to be at location 1. As can be seen from Figure 12, at 10*5 the mean magnitude and distance is about 6. 28 at 28 km. At 104 it is about half an order of magnitude lower and about 10 km further away. Also, given the acceptable probability level, the mean magnitude-and distance for each source varies depending upon the location of the site. Figure 13 shows site locations 1, 2, and 3. At a probability level of 104 , given that the site is in location 1, the mean magnitude associated with source A is 5.7 at 27 km and the mean magnitude associated with source Bis 7.0 at 248 km. If the site is at location 3, the mean magnitude associated with source A is 6.5 at 250 km and the mean magnitude associated with source Bis 6.0 at 27 km. As can be seen, depending upon site location, the controlling earthquake for source A can be either a magnitude 5.7 or a magnitude 6.5. The point to be made, is that the mean magnitude and distance is simply a statistical manifestation of the acceptable probability level which is the key parameter. Lastly, due to uncertainty about the attenuation model the acceptable ground motion used for deaggregation can not be duplicated at the site if the mean magnitude and distance determined from deaggregation is simply plugged into the attenuation model used in the hazard calculations. To arrive at the acceptable ground motion defined in step 2 using DG-1015, it will likely be necessary to apply an 'ad hoc' correction factor to the step 5 conversion of mean-magnitudes at some distance from the site to ground motions.at the site. See Figure 13 for examples. Therefore, given the above situation, it should be clear that the process of defining an acceptable ground motion at a site, deaggregc;lting this ground motion to a mean magnitude at some distance from the site, and then bringing the ground motion back to the site is unnecessary. Simply scaling either a standard spectral shape or a site specific shape based upon mean magnitudes and distances to the acceptable ground motion level is a defensible approach. This is consistent with DOE-STD-1024-92 (December, 1992). ouestion NUml)er 3 In determining the controlling earthquakes, should the median values of the seismic hazard analysis as described in Appendix c to Draft Regulatory Guide DG-1015 be used to the exclusion of other statistical measures, such as mean or 85th percentile? (The NRC staff has selected probability of exceedance levels associated with the median hazard analysis estimates as they provide more stable estimates of controlling earthquakes.) Answer to ouestion 3 Because either LLNL or EPRI may be used in the current version of 9

DG-1015, consistency of predicted results between the two methodologies is important. Based upon our analyses median results are most consistent between the two methodologies. With respect to establishing the reference probability for determining the SSE, the use of median hazard results is appropriate. However, if it can be shown that because of changes in the LLNL methodology that mean or other statistics are more consistent between LLNL and EPRI then that statistic should be used. It should be remembered that use of a standardized methodology (LLNL/EPRI computer code and protocol) in conjunction with standardized inputs (same experts for each site) results in consistent (within methodology) site-to-site comparisons (relative use of probabilities). Consistency between methodologies is currently accomplished through use of median hazard curves. Question Numl:>er 4 The proposed Appendix B to 10 CFR Part 100 states: "The annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median annual probability computed from the current [EFFECTIVE DATE OF THE REGULATION] population of nuclear power plants." This is a relative criterion without any specific numerical value of the annual probability of exceedance because of the current status of the probabilistic seismic hazard analysis. However, this require-ment ensures that the design levels at new sites will be comparable to those at many existing sites, particularly more recently licensed sites. Method-dependent annual probabilities or target levels (e.g., lE-4 for Lawrence Livermore National Laboratory or 3E-5 for the Electric Power Research Institute) are identified in the proposed regulatory guide. Sensitivity studies addressing the effects of different target probabilities are discussed in the Bernreuter to Murphy letter report. Comments are solicited as to (a) whether the above criterion, as stated, needs to be included in the regulation and (b) if not, should it be included in the regulation in a different form (e.g., a specific numerical value, a level other than the median annual probability computed for the current plants)? Answer to ouestion 4 As shown above, acceptable probabilities are consistent with a methodology (1. o x 104 for LLNL and 3. o x 10-5 for EPRI). Acceptability is defined by calibrating the methodology relative to a standard. In this context, the above definition simply defines the standard to calibrate a probabilistic methodology and therefore is sufficient. Implicit in the standard are the acceptable probabilities. However, what is most important is that the standard be stable. Use of the above definition does result in conservative acceptable probabilities, however there seems to be no quantitative basis for choosing the median probability level. Why not the mean or 15th percentile? To avoid this problem, we believe that the standard should be based upon a clearly acceptable deterministic standard. Because a 0.3g R.G. 1.60 spectrum exceeds the design level of any existing nuclear power plant in the EUS it 10

is by definition acceptable at each existing site. Therefore, the standard should state, "The annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the enveloping probability of exceeding a 0.3g R.G. 1.60 spectrum assumed at existing nuclear power plant sites [EFFECTIVE DATE OF THE REGULATION]. 11 It has been shown {10) that the acceptable probabilities using this standard are consistent with those defined by the current standard used in DG-1015. ouestion lfnpher s For the probabilistic analysis, how many controlling earthquakes should be generated to cover the frequency band of concern for nuclear power plants? (For the four trial plants used to develop the criteria presented in Draft Regulatory Guide DG-1015, the average of results for the 5 Hz and 10 Hz spectral velocities was used to establish the probability of exceedance level. Controlling earthquakes were evaluated for this frequency band, for the average of 1 and 2.5 Hz spectral responses, and for peak ground acceleration.) Answer to Question 5 The Safe Shutdown Earthquake Ground Motion response spectrum should be determined based on scaling an accepted response spectrum shape to the probabilistic seismic hazard analysis results for the average of 5 Hz and 10 Hz consistent with the reference probability level. A magnitude-distance pair can be determined for this spectral acceleration level. A site specific spectral shape can be developed based upon this mean magnitude and distance. This shape can be scaled to the acceptable ground motion value. The pref erred approach is to scale a standard shape to the acceptable ground motion value. Additional comments on the Determination of an Acceptable Design Value at :ruture Eastern u. s. sites The nuclear power industry is extremely aware of the competition presented by non-utility generators (independent power producers and co-generators). These non-utility generators can permit, build, license, and put on line a 600 megawatt plant in 4 to 5 years. For the nuclear industry to be competitive, the overall process to permit, license, build, and put a plant on line must also be in a similar time period. The following approach will result in acceptable seismic design values at EUS sites and will satisfy the need for a timely licensing process. A Recommended Seismic Siting-Approach (SSA) - The Philosophy of this Approach Should Replace the current DG-1015 The essence of this alternative seismic siting approach is the calibration of a seismic hazard methodology relative to a conservative standard ( o. 3g R. G. 1. 6 o spectrum) to determine acceptable probabilities, and then the use of these probabilities 11

in an internally consistent manner. The details of the SSA are documented in ( 9, 10) . Figure 8 shows there is reasonable consistency between the LLNL and EPRI results in a relative sense but not in an absolute sense. Given this, it is only a matter of defining acceptable probabilities consistent with a methodology to determine acceptable site design levels. The precepts of the SSA are as follows:

1. A 0.3g R.G. 1.60 spectrum assumed at existing plants is adequately conservative to establish stable reference probability levels. Tµe enveloping probabilities define acceptable probabilities (10).

2. Existing LLNL .sng EPRI seismic sources and parameters are an acceptable basis for assessing a site's SSE ground motion in the EUS. These inputs reflect the judgements of nationally recognized experts and form the basis for determining the acceptable ground motion at the site.

3. Median hazard curves are most consistent between methodologies and therefore are used to determine acceptable ground motion values.

4. Both LLNL and EPRI methodologies are used to determine acceptable ground motions - the higher of the two results are used.

5. A standard spectral shape is scaled to the acceptable ground motion level (consistent with DOE-STD-1024-92).

6. Close-in site-specific geological, seismological and geophysical investigations must be performed. Results of these investigations are used to confirm the suitability of the site * .

7. The SSE is considered invalid if evidence of foundation material inadequacy or local capable faulting is determined.

8. Implicit is the assumption that every 10 years the data base for each of these methodologies will be updated.

9. EXplicit is the condition that a 0.3g R.G. 1.60 spectrum is acceptable at any existing EUS nuclear power plant site.

This approach results in an acceptable site design level of less than O. Jg at typical EUS sites ( excepting the areas around New Madrid, Missouri and Charleston, South Carolina). A standardized plant design level of 0.3g is certainly more than acceptable for these sites. Furthermore, using this approach, the probability of exceeding the acceptable site design level will be the same, from site-to-site, while the probability of exceeding the actual plant design value will always be equal to or less than the probability of exceeding the acceptable site design value. 12

As stated earlier, "consistent" is defined in terms of probabilities specific to a given methodology. Using these acceptable probabilities, standardized probabilistic methods (such as LLNL, EPRI, USGS, or the results of a resolution between LLNL and EPRI) can be used to determine acceptable site design levels that will be consistent in terms of probability of exceedance from location to location. The philosophy of this approach is similar to that of Short, et al (11), except that they advocate the use of the mean probability of exceeding current SSE values. Based upon arguments developed in ( 10) , we conclude that the enveloping probabilities of exceeding the 0.3g spectrum at existing EUS sites should be used to define acceptable probabilities for future sites. Using the LLNL and EPRI hazard results for existing sites and these acceptable probabilities at the PGA, 25, 10, 5, 2.5, and 1 bz, acceptable site seismic design spectra for future plants at existing sites were determined. Figure 14 shows these proposed spectra relative to the 0.3g R.G. 1.60 spectrum using LLNL hazard results. Similar results, but not exactly the same results, would be obtained if the EPRI hazard results were used. As can be seen, Figure 14 looks very similar to Figure 15 which is the current design levels at existing sites; however, the significant difference is that all of these spectra have exactly the same probability of exceedance which is defined by the enveloping probability of exceeding a 0.3g spectrum at existing sites. In other words, the probability of exceeding the spectra would be consistent across all sites. Because the results are not exactly the same, and because the cost of these probabilistic analyses are relatively low, it is recommended that .l2.Q.th the LLNL and EPRI analyses be performed and the higher of the two spectra be used to define the site seismic design level. An important outcome of this overall approach is that it assures that a 0.3g standardized design is acceptable at any existing EUS site regardless of seismic hazard methodology used. Furthermore, future sites, similar in hazard to existing NPP sites, are assured that their site seismic design level will be a.Jg or less. For those locations where the site seismic design level is less than 0.15g, a minimum site design SSE of 0.15g is assigned. A minimum design value of 0.15g is chosen because it is about the average design value for the current population of NPPs. To be consistent with the current DG-1015 approach, the acceptable ground motion can be determined from the average of the 5 and 10 hz hazard curves. However, a standard spectral shape, such as a R.G. 1.60 modified at high frequencies, is simply scaled to the acceptable ground motion value. From (4) the following is excerpted. 'In most of the United states, and particularly east of the Rocky Mountains, the SSE is usually determined on a so-called tectonic province approach, because historical earthquakes are relatively infrequent and few can be clearly correlated with geologic structures.' The LLNL ~d EPRI methodologies can be viewed as creating pseudo-tectonic provinces that in a reproducible fashion define the SSE for a site. Consistent with the Appendix A philosophy and past licensing practice, it is extremely important that well defined, close in. seismological, and geological evidence around the proposed site be gathered to confirm the suitability of the site 13

given the SSE. The SSE spectrum is considered invalid only if evidence of foundation material inadequacy or local faulting is determined. This is consistent with the present Appendix A philosophy. It is our belief, that by adopting a conservative standard to determine acceptable probabilities, and by using .Qgj;;h the LLNL and EPRI methodologies (or the resolution methodology), conservative SSE values at a future site are insured. Site-specific information unique to the site (faulting etc.), should be assessed consistent with past licensing decisions - that is, deterministically. Figure 16 is a flow diagram of our siting approach. This approach applies to only EUS sites. Until a comparable data base for the WUS is developed, deterministic (Appendix A) methods are assumed to apply. This is consistent with DG-1015 philosophy. This alternative siting approach has been developed to stabilize the licensing and regulatory process. This approach is based upon our licensing experience with all external events and in particular with our experience associated with the determination and licensing of seismic design bases for existing NPPs. This is primarily a probabilistic process that is to be complimented with detailed and meaningful deterministic evaluations at the site. Results of the deterministic evaluations are used only to invalidate the SSE. The challenge to the technical community is to review the currently required site/site area studies and clearly define criteria that confirm or deny site suitability. our approach results in an acceptable site design level for future NPPs of less than 0.3g at typical EUS sites (regardless of methodology (LLNL or EPRI)) and a default standardized plant design level of 0.3g. If this approach results in a site seismic design level of less than 0.15g, a minimum site design SSE of 0.15g is assigned. Based on our approach the site design level is calculated using both the LLNL and EPRI methodologies with the higher of the two estimates defining the acceptable site seismic design level. This conservatism is adopted to avoid unnecessary debate concerning choice of methodology. This approach can be readily incorporated into the framework of regulatory guides. For future sites, the LLNL and EPRI methodologies should be assumed as pseudo-tectonic province maps that define acceptable site seismic design levels for essentially all of the EUS. Site-specific geologic and geotechnical analyses would confirm the suitability of the site. Utilities that choose to bound the acceptable site value with a 0.3g standardized design may find it beneficial to use the acceptable site SSE when performing liquefaction and other geotechnical analyses. 14

References

1. "Probabilistic Seismic Hazard Analysis," National Academy Press, Washington, D.C. 1988

2. Electric Power Research Institute, 1989, "Probabilistic Seismic Hazard Evaluation at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Earthquake Issue," prepared by R. K. McGuire, et al, Electric Power Research Institute Report NP-6395-D.

3. Lawrence Livermore National Laboratories, 1989, "Seismic Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky Mountains," prepared by D. L. Bernreuter, et al, USNRC, NUREG/CR-5250.

4. Hatheway, A. w., and McClure, C.R., Jr, 1979, "Geology in the Siting of Nuclear Power Plants," Geological Society of America, Reviews in Engineering Geology, Volume IV *

5. Minogue, R. B., SECY-79-300, "Identification of Issues Pertaining to Seismic and Geologic Siting Regulation, Policy, and Practice for Nuclear Power Plants," USNRC, April 27 I 1979 *

6. US Nuclear Regulatory commission, 1991, "Procedural and su_.~mittal Guidance for the Individual Plant Examinations of EXtt'!1:'nE=l Events (IPEEE) for Severe Accident Vulnerabilities, 11 USNRC NUREG-1407, Page 103.

7. Lawrence Livermore National Laboratories, 1985, "An Approach to the Quantification of Seismic Margins in Nuclear Power Plants,"

prepared by R. J. Budnitz, P. J. Amico, c. A. Cornell, w. J.

     ">11, and R. P. Kennedy, . USNRC NUREG/CR-4334.

8. Davis, J. F. et al, 1979, "The state-Federal Partnership in the Siting of Nuclear Power Plants", Geological Society of America, Reviews in Engineering Geology, Volume IV, page 47.

9. O'Hara, T., and Jacobson, J. P., 1990, "Seismic Hazard Analyses

     - A Utility Perspective," Nuclear Engineering and Design, Volume 123, Pages 111 - 122.

10. O'Hara, T., and Jacobson,J. P., Briggs, W.J. 1992, "Standardized Seismic Design (SSD} for Nuclear Power Plants - A Utility Perspective," Nuclear Safety, Volume 33-4.

11. Short, S. A., Murray, R. C., and Hill, J. R., 1990, "Deterministic Seismic Design and Evaluation Criteria to Meet Probabilistic Performance Goals,"Third Symposium on Current Issues Related to Nuclear Power Plant Structures, Equipment, and Piping, Page 12.

1

PROBABILISTIC SITE DETERMINISTIC ANALYSIS (PA) ANALYSIS (DA) Geological, Seismological and Geophysical lnvestlgatlons I Conduct an EPRI or LLNL Identify Seismic Hazard Assessment Seismic Sources

• Compare to Operating Plants to Set Probability of Exceedance Level Determine Deterministic Source Earthquakes for Each Source Determine Controlling Determine Controlling Earthquakes (CEs) Ms & Ds Earthquakes (CEs) Ms & Ds

                                        ,I Compare CEs Derived From PA and DA I

Develop SSE Ground Motion (GM) or Compare with CE GMs Figure 1. Flow chart for determination of the SSE In the eastern United States. (DG-1015, pg 11)

0 50Mlln m:s50 K1lometer1 0 N .'t'.

---

- - - - -i..,

                      '\

PA. I!!, White Mountain intrusives 0 Cape Ann earthquakes A Location of nuclear units

1. Millstone 1, 2, 3

2. Plymouth 1, 2

3. Seabrook 1, 2

4. Indian Point 1, 2, 3 Figure 2. Selected nuclear power stations and pertinent geologic features. Heavy lines indicate boundaries of the New England-Piedmontf Southern Platform, and Coastal Plain tectonic provinces advocated by NRC testimony in the Indian Point Appeals Board Proceedings (from reference 1).

r:

r: 0)

• o C

                                                                  ~ 0
                                                                          ....J

r:

Figure 3, LLNL host zones forthe Seabrook site <LLNL, 1989).

z i

0') j

z I

                                                           -:II I

Figure 4. LLNL host zones for the Vogtle site (LLNL, 1989).

7.5 -------------------'9'----- LLNL Upper Bound Magnitude Estimates for the Seabrook NPP o Host Zones 7 -- - - - - - - - - - - - - - - - -A- - - - - - - - E)- - - - - - - - *

                     = mean of di.stribution      A                 e0
     ~ 6.6 - - - - - - - - - - - - - - - ~ - - - - - - - -      ) - - - - - - - - *
      ~                                                               0
     ~                              B              !                 o g>           Seabrook                                           o
     ~     6         SSE t:,,.
                                     --------A---------.-----;_                     __ _

t:,,. 6.6 - - - - - - - - - G - - - - - - - - - - - - - - - - - - - - - - - - - -

• s------*-_____, _____,________.

0 1 Best L.::.. 2 3 4 Upper Bound Magnitude Estimate Figure 5. Distribution of host zone upper bound magnitudes at Seabrook. 7.6 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _-'1,,,,_ _ _ __ LLNL Upper Bound Magnitw;Ie 0 Estimates for the Vogtle N'Pr Host Zones ** 7 ,..._ - - - - - - - - - - - - - - - -b- - - - - - - -e- - - - - - - -

                    = mean of distribution         A 0
     ~ 6.5 "' - - - - - - - - - - - - - - - - - A - - - - - - - - - - - - - - - - -
      ~                                             L                 0
      ~                             0

• 8 C,

     ~

a, 6 .... - - - - - - - - ------- 0

                                                 -A - - - - - - - - - - - - - - - - -

0

                                                    ~                 0 6.5 ... - - - - - - - - - - - - - - - - - A - - - - - - - - - - - - - - - - -

Vogtle 6 .___ _ _ _-i..... 0 i-o-------*-----*------... 1 2 3 4 Upper Bound Magnitude Estimate Figure 6. Distribution of host zone upper bound magnitudes at Vogtle.

SEISMIC DESIGN LEVEL (SSE) AT eus NPP SITES

   -C) 0.2
   <(

CJ

0. 0.15 l-+HH----t-<il--9-f>-+--~-...+......-...-+~+-t+--+-ff--+---t-f----i 0.1 0.05 ........_..........._.__..__..._.__,__,_..................__.__........_..__._..................___.__._.._.__.__.__..__,__.__........_........

0 10 20 30 40 50 60 70 SITE Figure 7. Distribution of seismic design levels (SSE) for EUS NPP sites. ANNUAL PROBABILITY OF EXCEEDING SSE LLNL/EPRI MEDIAN HAZARD RESULTS

PROBABILITY OF EXCEEDING SSE SPECTRA FOR ALL EASTERN U.S. SITES (LLNL 5gx MEDIAN, 5% DAMPING) 0.01 - - - - - - - - - - - - - - - - - - - 0.001

 ~

t:

 ....:I 1--1 p:i        0.0001 p:i 0
 ~

p...

 ....:I
 ~        0.00001 z

z 0.000001

                                    -------- LLNL MEDIAN 0.0000001 ______________....._.._,___..i.-_ _ _ _ _ _ _ _ _ _.._.__

1 10 100 FREQUENCY (HZ) Figure 9. Probability of exceeding SSE spectra for EUS NPPs.

Comparison of Predicted Target Ground Motion Values from LLNL and EPRI Studies (Cl C oJ

                                                                             !                       LLNL    EPRI        I
     .0                .
                ,...........                                                 f\                    ~--------~ Target values above this line t)                                                       il I   I exceed 0.3g A.G. 1.60 spectra.
      ~           Q)                                                       :   I mo O"

CD CD Je

                                *************************--**-***--**---i*----*******--*-*******-----**-*************--*---------***-******************--**-*********-********-
                                                                          ! l I

I

• C.

g

    ¥             E t)

I I

• I I

r CD r c O" I

                                                                                              ~

z-, ~10 ~ rg *5 P> :,

, C.

0. 3

               -0 Cl) mo

-a

n 0 ->ctS **

en ~- pl

, I...

t5 ..* 3 ~ Cl) a.

                                --*~--- .'~--**...

er en

,- a en ...

~ ~

a. :,~

CD (Q en en 0 C

+ r+-* 0 10 20 30 40 50 60 en

• en <D Site Number

E.U.S SEISMIC HAZARD CHARACTERIZATION LOWER MAGNITUDE Of INTEGRATION IS 5.0 BEST ESTIMATE

                -1                        roR THE SEISMICITY EXPERTS 10
                -2 10 "Q:

c( \ l;J Q:

                -3 i..i Q. 10 Lo.I u
          ?iQ l;J w     -4 u    10 X

w I.&. 0

          !:    -s
          ..J  10 CD CD                   '-

0 'C,, er Q.

                -6 10
                -7 10
                                      ..,   .... in   IQ    ,.. CD en ACCELERATION CM/SEC**2 FARLEY Figure 11 . LLNL best estimate results by expert for the Farley site.

Mean Magnitude and Distance For a NPP Site

!] in New England ca C

CD I\) M = 6.28

0 ..

p)

, ~ ..

a. ~ . -- 40

-a 0 o en Q) '***.......,&. ~, g~ Q)

             'U 6                                                                                                                                  u O" -a         ::J
=:    O"     :!::                                                                                                                                  C C                                                                                                                                    co
                                                                                                                                                   +-'
<D <D

0) en

~     CD      co                                                                                        A= 28 km                                30 0

- :::, ~

3 ----------A .. ________8------------- C

      <D Pl C

ro - ---- ..... co

, Q)

Q) 3 ~

             ~
     ~                                                                                               - - -- --            - - -----             20 C              M,R"' 5.3 at 46 km a.
      <D M,R    = 5.84 at 35 km        M,R "' 6.28 at 28 km 3                     1.0e-3                   1.0e-4                       1.0e-5 CD p) 0..

5 ~~~--~~~-~~~~~~~~-~~~-~~~~ I I 10 cii"

      ..+

0 100 200 300 400 500 600 700 800 900 1,000 p) 0 Acceleration (cm/sec**2)

      <D

z:

NORTHERN APPALACHIANS ZONES iBr----------...,....,r:---i . .. 0

                'Z

• ~-----

          'Z 0

Relationship between deaggregation level, (M,R), attenuation, and M,R by source. Attenuation Model Ln(AP) = 2.55 + 1.0*M - 0.0046*R - 1.0*ln(R) Deaggregation Given M,R Source Site PGA at 1 E-4 M,R PGA at Site A B M,R M,R 1 0.195g 5.7@27 0.13g 5.7@27 7.0@248 2 0.265g 5.9@25 0.17g 5.8@29 5.9@25 3 0.292g 6.0@27 0.17g 6.5@250 6.0@27 Figure 13. Seismic source zones and mean magnitude and distance results at various locations.

FUTURE SSE SPECTRA FOR EASTERN U.S. SITES BASED ON LLNL METHODOLOGY ________ 0.3g R.G. 1.60 SPECTRUM

• 04 _ _ _ _ _ _ _ _ _.........._ _ _ _ _ _ _........__

1 10 FREQUENCY (HZ) 100 Figure 14. Predicted SSE spectra at existing sites based on SSD approach. SSE SPECTRA AT ALL EASTERN U.S. SITES (5% DAMPING) 0 . 1 - - - - - - - - - - - - - - - -........... 1 10 100 F1IBQUENCY (HZ) Figure 15. SSE spectra for all EUS NPP's, and R.G. 1.60 spectrum.

EUS Site Deterministic Probabilistic I Both LLNL and EPRI methodolog/ea used and the higher of the two tu1ed at the alte. CONDUCT S1'. GEOLOGI~ SEISMOLOGICAL -~--- Ampllflcatton - Ll.NL AND EPRI HAZARD AND GEOPHYSICAL (Son sites only) ASSESSMENT FOR SITE IN\/ESTIGA110NS (EMPHASIS ON SITE AREA) BASED UPON *- - ,_ PR06ABIUTIE8 ut:, .......... HIGHEST ACCEPTABLE SIT : GROUND MOTION VALUE

• SITE SPECIRC SPECTRUM I

BA.SEO ON MEAN MAGNffiJOE ANO DISTANCE DERIVE THE SPECTRAL SHAPE I SCALE SPECTRAL SHAPE SCALE MOOIFIED R.03. 1 110

                                                 .TO ACCEPTABLE GAOUNC                                  TO ACCEPTABLE MOTION LEVEL                                       SITE GROUND MOTION I

I SSE I I SSE I Results can only Invalidate SSE given evidence of -- Confirm foundation matertal Inadequacy SSE based on site area or local capable faultlng geotechnical Investigations I SSE Figure 16. Proposed flow chart for determination of the SSE in the eastern U.S.

Suggested Changes Either Highlighted or Bold October 1992 Division 1 Draft DG-1016

Contact:

R. M. Kenneally (301) 492-3893 DRAFT REGULATORY GUIDE DG-1016

• 1 (Modifications by Yankee Atomic and Northeast Utilities)

NUCLEAR POWER PLANT INSTRUMENTATION FOR EARTHQUAKES A. INTRODUCTION 2 3 In 10 CFR Part 20, "Standards for Protection Against Radiation," 4 licensees are required to make every reasonable effort to maintain radiation 5 exposures as low as is reasonably achievable. Paragraph (c) of S 50.36, 6 "Technical Specifications," to 10 CFR Part ~O, "Domestic Licensing of Pro-7 duction and Utilization Facilities,* requires the technical specifications 8 of a facility to include surveillance requirements to ensure that the neces- -~ 9 sary quality of systems and components is maintained, that facility opera-tion will be within safety limits, and that the limiting conditions of operation will be met. Paragraph IV(a)(4) of Proposed Appendix s, "Earth-12 quake Engineering Criteria for Nuclear Power Plants,* to 10 CFR Part 50 13 would require that suitable instrumentation be provided so that the seismic 14 response of nuclear power plant features.important to safety can be evalu-15 ated promptly. Paragraph IV(a)(J) of Proposed Appendix S to 10 CPR Part 50 16 would require shutdown of the nuclear power plant if vibratory ground motion 17 exceeding that of the operating basis earthquake ground motion (OBE) 18 occurs.* 19 20 *Guidance is being ~eveloped in Draft Regulatory Guide DG-1017, "Pre-21 Earthquake Planning and Immediate Nuclear Power Plant Operator Post-22 Earthquake Actions," on plant shutdown criteria. 1

1 This guide is being developed to describe seismic instrumentation f;§#?#i.~ 2 n~¢¥'$*;:::::,SfM~:f:pX,1nt*d=~J=..:tvJt* acceptable to the NRC staff for satisfying the 3 requirements of Parts 20 and 50 and the Proposed Appendix S to Part SO. 4 Any information collection activities mentioned in this draft regulatory 5 guide are contained as requirements in the proposed amendments to 10 CFR 6 Part 50 that would provide the regulatory basis for this guide. The proposed 7 amendments have been submitted to the Office of Management and Budget for 8 clearance that may be appropriate under the Paperwork Reduction Act. Such 9 clearance, if obtained, would also apply to any information collection 10 activities mentioned in this guide. 11 12 B. DISCUSSION 13 When an earthquake occurs, it is important to t:1¥:!fii#si!i:i'Ht.EB+in@I§ assess the effects of the earthquake at the nuclear power plant. This 16 assessment includes both a plant walkdown and the evaluation of seismic 17 instrumentation. lol i d-state digital time-history accelerographs installed 18 at appropriate locations will provide time-history data on the seismic 19 response of the free-field, containment structure, and other Category I 20 structures. The instrumentation should be located so that a comparison and 21 evaluation of such response may be made with the design basis and so that 22 occupational radiation exposures ii'!ss:!¥1ielHIIP:ifiifii!t:I:eiisi:¥£§i are 23 maintained as low as reasonably achievable (ALARA). 24 Instrumentation located on the containment foundation is used to 25 determine if the OBE has been exceeded (see Draft Regulatory Guide DG-1017). 26 Foundation-level instrumentation provides data on the actual seismic input to the containment and other buildings and can be used to quantify differences between the vibratory ground motion at the free-field and foundation level. 29 Instrumentation is not located on equipment, piping, or supports since experi-30 ence has shown that data obtained at these locations are obscured by vibratory 31 motion associated with normal plant operation. 32 The guidance being developed in Draft Regulatory Guide DG-1017 is based 33 on the assumption that the nuclear power plant has operable seismic instrumen-34 tation, including the equipment and software required to process the data 35 within 4 hours after an earthquake. This is necessary t.qn~--, --~:u.i*dff.:tpt;l!fi,~ 36 JJ.J@.l(t4§~ 1;Wg~qjt~@):f~.:::;DqdJ*#.j.~A#.;iQA:::::w$.l.1fal.~:Jlffl.'-"- by comparing the 37 recorded data against OBE exceedance criteriaJ@JJffl the results of the jJ!jifi 38 walkdown inspections that take place within 8 hours of the event. 39 It may not be necessary for identical nuclear power units on a given site 40 to each be provided with seismic instrumentation if e sse nt ially the same 41 seismic response at each of the units is expected from a given earthquake . 42 An evaluation of seismic instrumentation operational experience noted 43 that instruments have been out of service during plant shutdown and sometimes 44 during plant operation. The instrumentation system should be operable at all 2

1 times. If the seismic instrumentation ri'i~'iJU~jJ#:j#\J§~lfi.ffiiH[{lf,t.i#:@$.ni:e.'tp.ff#:~;N~ffl-2 is inoperable, the guidelines in Appendix I to Draft Regulatory Guide DG-1017 3 would be used to determine whether the OBE has been exceeded. 4 Instrumentation characteristics, installation, activation, remote indica-5 tion, and maintenance are described in this guide to help ensure (1) that the 6 data provided are comparable with the data used in the design of the nuclear 7 power plant, (2) that exceedence of the OBE can be determined, and (3) that 8 the equipment will perform as required. 9 The Appendix to this guide provides def i n i tions to be used with this 10 guidance. 11 12 C. REGULATORY POSITION 13 The type, locations, operability, characteristics, installation, actuation, remote indication, and maintenance of seismic instrumentation 16 described below are acceptable to the NRC staff for satisfying the require-17 ments in 10 CFR Part 20, 10 CFR 50.36(c), and Paragraph IV(a)(4) of Proposed 18 Appendix s to 10 CFR Part 50 for ensuring the safety of nuclear power plants. 19 20 1. SEISMIC INSTRUMENTATION TYPE AND LOCATION 21 22 1.:..l f olid-state digital instrumentation that will enable the processing 23 of data at the plant site within 4 hours of the seismic event should be used. 24 25 ~ A triaxial time-history accelerograph should be provided at each of 26 the following locations: 1. 29 30 2. Containment foundation. 31 32 3. ~ii

                      ,:*?""""' '*

elevation (excluding the foundation) on a structure 33 internal to the containment. 34 35 4. bhe *independent Category I structure foundation (for instance, 36 the diesel generator building 9:~ the auxiliary building) where 37 the response is different from that of the containment 38 structure. 39 40 41 1.3 The specific locations for instrumentation should be determined by 42 the nuclear plant designer to obtain the most pertinent information consistent 43 with maintaining occupational radiation exposures ALARA for the location, 44 installation, and maintenance of seismic instrumentation. In general: 3

1 .!.:.1..:..1 A design review of the location, installation, and 2 maintenance of proposed instrumentation for maintaining exposures ALARA should 3 be performed by the facility in the planning stage in accordance with 4 Regulatory Guide 8.8, "Information Relevant to Ensuring that Occupational 5 Radiation Exposures at Nuclear Power Stations Will Be As Low As Is Reasonably 6 Achievable. " 7 8 1.3.2 Instrumentation should be placed in a location with as low 9 a dose rate as is practical, consistent with other requirements. 10 11 1.3.3 Instruments should be selected to require minimal 12 maintenance and in-service inspection, as well as minimal time and numbers of 13 personnel to conduct installation and maintenance. J&-i!l:1N=dWb)1rti#.ntl=i.#.:X.4i.t!.!:>>l'*-#~ttM#:!Alt.4=,:,,§,tWrt.~~=J:\,~f4.:@iJ*-4lJ:pg*tJti:f.='flf.§. 16 t:b*Jf$.fflib.t\P)?-P:ta..il)m)iW!-#ttt~t¥lfr§,-:fp.f:mtn*::::::m;p.µ~@t:::~w;@~P~+~n9.tJtJ.ffin~t.,(t.i.9:t,t::t11t,-,: 17 :u*k**d.$ii'.tiR:tw.$tl1,::,,~e~~#.,;¥.,%]\:\ifµ~~*<<!Wf"~§µ:#,..'f:tJ@:$.W.4.:~=µg$?,,i,J.~4.:ff9.,9~~>>1:1:tw$+:J.::::,:p.* 18 iq~i.nt'fl=\Pfi~t:a:tl,\'¥:¥£t.ijg=::::;t~,~~t:t1f;e.:I~h=:iumf'-Plt:i1n=,~:fip.~,14,,,,,1~:=::=Pl4'd:1a::,:=,~tt,;t~(~t;tP.p,fa,::pa:ae.a 19 P.n.i.hlPP#P:lW(f.4~.,}=~fiV!.f.~$$E#.niFe~t~~#,,~ffi!P~tn:tt!'.9.#l:$~¥~P.+./;i.}tM4'.%#.h~ttlti~e#.;t@!- 20 :t-oW1tii'-t:o.pN#W}tJ~~%*qfQ;,t#l::m~wr:§~::,:::tt.~p=<<JH~~W$:;t.§~!:#Jµ.g:::,:i.:t:;k~rir4.,~g~w+~w.\'-Y::J,fr;pm 21 i!!!R~S!iI%R£'{8!1fal8illlR$!5i:ti!R!I\J?:!!t£biMf$9:R!\!\!ntiM\fM\~tff:efiMti=Wtl\llP£i!!ltlei 22 #.9§J#,;,~J.#t.HW*-t@aiiffiP¥i#'.'t.i.J;i¥\tW.ffe.tf§!~llii&.#.i;.W:#HWfi+#tWMt.it.)H===#.ffe.#1l'i#,fiA=t.;~f===#!F#.~~KJ#~ 23 #l!llM!!Ri!tK!!ffiIRi!B~%4!l!~fMUlllJ+g~f9JfJ@1+~iHflM¥:l=~ieflllfi!:l\il#:S~!4llffi#:IA!!JIIl!!I: 24 eJji)jg;iyj:;.:t,q\tlJ\~!fQ.\,MHi"=ff.PP.#:P:!Wi.~U;*#:~=!~tV£J.t,:$#.Ji.#~#.f.=lil'9¢~¥.>>-P-!i!OJt+!:ttl# 25 ...#t:JdtD.1#:t=#*~*is'-P.tttJd;jJ#N.ik!,f~#+.?P,tJ:ii#!if#p#.J~=+~#.)tp.#')d;:t;J,g==.,,y,.,mt.4t.1-t.#t#.9 26 t.tt.\'t.V.l1'$9#.@tn4WiM=f:AJ.¥.ffl1/4J.t.'tM,#.1j.\,,,:~#.:i,i'.ffi'tf,ffl2Y.~l~>>Pij:@t.~fH,1~tl~$#.!,¥&¥:=,=,=iti##.*'#:#f.jM 29 30 2. INSTRUMENTATION AT MULTI-UNIT SITES 31 32 Instrumentation in addition to that installed for a single unit will not 33 be required if essentially the same seismic response is expected at the other 34 units based on the seismic analysis used in the seismic design of the plant. 35 However, if there are separate control rooms, annunciation should be provided 36 to both control rooms as specified in Regulatory Position 7. 37 38 3. SEISMIC INSTRUMENTATION OPERABILITY 39 40 The seismic instrumentation should operate during all modes of plant 41 operation, including periods of plant shutdown. The maintenance and repair 42 procedures should provide for keeping the maximum number of instruments in 43 service during plant operation and shutdown. 44 4

1 4. INSTRUMENTATION CHARACTERISTICS 2 3 4.1 The design should include provisions for in-service testing. The 4 instruments should be capable of periodic channel checks during normal plant 5 operation. 6 7 4.2 The instruments should have the capability for in-place functional 8 testing. 9 10 4. 3 I nstrumentation ~ttlttiid:~m~q#~tft&$.at.,);tt,i.W@t.f:li~¢.~~,s.;p?-l.:~t:~f$.a~ should 11 contain prov1.sions for ~'!~lWF@E§.#SWA9.#e.E}i~H.tf4P:9:~,~!-~!!@liliP.¢.~JAi?:ttt,~~Sl$ib'- 12 'ffi.AfHfiffi@@!#.\t@.lWlii!b.'-#,lA'tf.#tj~:fi4,~/~jftj~ an external remote alarm to indicate 13 actuation. 4.4 The pre-event memory of the instrumentation should be sufficient to 16 record the onset of the earthquake; for example, it should have the ability to 17 record the 3 seconds prior to seismic-trigger actuation. It should operate 18 continuously during the period in which the earthquake exceeds the seismic-19 trigger threshold and for a minimum of 5 seconds beyond the last seismic-20 trigger signal. The instrumentation should be capable of a minimum of 25 21 minutes of continuous recording ett:=~l=Eii!R!ll=e!lii!ssi!lfi$.il!fiiil!'!!a!=fl:9 22 m£nhte.iJJ:dtHMni.i'fifl,ffiat.tan. 23 24 4.5 Acceleration Sensors 25 26 4.5.1. The dynamic range should be 1000:1 zero to peak, for 27 29 30 31 example, O.OOlg to 1.0g. 4.5.2. equivalent demonstrated to be adequate by computational techniques applied to the resultant accelerogram. The frequency range should be 0.20 Hz to 50 Hz, or an 32 33 4.6 Recorder 34 35 4.6.1. The sample rate should be at least 200 samples per second 36 t 3wj1¢.b.fi#.lif#.ffitJ$.,6J;*; Nf.:t.dg$i.tm,. 37 38 4.6.2. The bandwidth should be at least from 0.20 Hz to 50 Hz. 39 40 4.6.3. The dynamic range should be 1000:1. 41 42 4.7 Seismic Trigger. The actuat i ng level should be adjustable and 43 W:l.it J.t*-1qm#.n@t~#.:~ngg of O. 005g to O. 02g. 44 5

1 Efa=:{~fiffiilffipym,,ntft,:::#b99.:ta,,,,:ttav.n+a.Jif:ff!ie.rd:tJ:i,#te.,;y\:eapac.it.y:=,,f:ab.:::::@.!i\%t;t 2 *~*Wt.uW#ift\It?$,~\=i'ffll$¥:*i=M.gy9,:g*,-n*~w:,=;w.4;~aQ.ij;e,r1tt~~t>>Pffli#it,=~e~~e.n.::q;i,~t~r,: 3 mali\~¥ln.ajlc.E,¥izjfi@SW.===@p'¥~Pia=tsm 4 5 5. INSTRUMENTATION INSTALLATION 6 7 5.1 The instrumentation should be designed and installed so that the 8 mounting is rigid. 9 10 ~ The instrumentation should be oriented so that the horizontal axes 11 are parallel to the orthogonal horizontal axes assumed in the seismic 12 analysis

• wu,e.1w1t\i.i.ill.W.4.!@Pim:::t~~h.*:~,,::;.p.@~~P.#.l:j::::~f@.Y.:t41:§~=i:,~9~m~nt:w.,fa 13 5.3 Protection against accidental impacts should be provided.

16 6. INSTRUMENTATION ACTUATION 17 18 6.1 Both vertical and horizontal input vibratory ground motion should 19 actuate the same time-history accelerograph. One or more seismic triggers may 20 be used to accomplish this. 21 22 6.2 Spurious triggering should be avoided. 23 24 6.3 The seismic trigger mechanisms of the time-history accelerograph 25 should be set for a threshold ground acceleration of not more than 0.02g. 26 ea 27 29

7. REMOTE INDICATION Activation of the free-field or any foundation-level time-history 30 accelerograph should be annunciated in the control room. If there are two or 31 more control rooms at the site, annunciation should be provided to each 32 control room.

33 34 8. MAINTENANCE 35 36 8.1 The purpose of the maintenance program is to ensure that the 37 equipment will perform as required. As stated in Regulatory Position 3, the 38 maintenance and repair procedures should provide for keeping the maximum 39 number of instruments in service during plant operation and shutdown. 40 41 8.2 Systems are to be given channel checks every 2 weeks for the first 3 42 months of service after startup. Failures of devices normally occur during 43 initial operation. After the initial 3-month period and 3 consecutive 44 successful checks, monthly channel checks are sufficient. The monthly channel 6

1 check is to include checking the batteries. The channel functional test 2 should be performed every 6 months. Channel calibration should be performed 3 during refueling. 4 5 D. IMPLEMENTATION 6 7 The purpose of this section is to provide guidance to applicants and 8 licensees regarding the NRC staff's plans for using this regulatory guide. 9 This proposed revision has been released to encourage public 10 participation in its development. Except in those cases in which the 11 applicant proposes an acceptable alternative method for complying with the 12 specified portions of the Commission's regulations, the method to be described 13 in the active guide reflecting public comments will be used in the evaluation of applications for construction permits, operating licenses, combined licenses, or design certification submitted after the implementation date to 16 be specified in the active guide. This guide would not be used in the 17 evaluation of an application for an operating license submitted after the 18 implementation date to be specified in the active guide if the construction 19 permit was issued prior to that date. 20 21 22 23 24 25 26 -27 7

1 APPENDIX 2 DEFINITIONS 3 4 Acceleration Sensor. An instrument capable of sensing absolute acceleration 5 and transmitting the data to a recorder. 6 7 Channel Calibration (Primary Calibration). The determination and adjustment, 8 if required, of an instrument, sensor, or system such that it responds within 9 a specific range and accuracy to an acceleration, velocity, or displacement 10 input, as applicable, traceable to the National Institute of Standards and 11 Technology (NIST), or responds to an acceptable physical constant. 12 13 Channel Check. The qualitative verification of the functional status of the 14 instrument sensor. This check is an "in-situ" test and may be the same as a 15 channel functional test. 16 17 Channel Functional Test (Secondary Calibration). The determination without 18 adjustment that an instrument, sensor, or system responds to a known input, 19 not necessarily traced to the National Institute of Standards and Technology 20 (NIST), of such character that it will verify the instrument, sensor, or 21 system is functioning in a manner that can be calibrated. 22 23 Containment - See Primary Containment and Secondary Containment. 24 25 Operating Basis Earthquake Gro~~j Motion (OBE). The vibratory ground motion 26 ~iiutJ.Br~Jlf.!f~ij.J."th*~"4£a;.,tfiW:ffl-lfffl~g,;gn:tM#.J!~t~n~,-i£JbS~i't.ti&~l~Jli!tlWiJI 27 tiif:{lffl:'fs@il@mijjKJl{H~-.:~-::.:,$!ffi~¥. 28 - 29 30 31 32 33 Primary Containmei:;. The principal structure of a unit that acts as the barrier, afte~ the fuel cladding and reactor pressure boundary, to control the release c _ radioactive material. The primary containment includes (1) the cont~.* nment structure and its access openings, penetrations, and appurte-

    ;,ances, ( 2) the valves, pipes, closed systems, and other components used to 34 i solate the containment atmosphere from the environment, and (3) those systems 35 or portions of systems that, by their system functions, extend the containment 36 structure boundary (e.g., the connecting steam and feedwater piping) and 37 provide effective isolation.

38 39 Recorder. An instrument capable of simultaneously recording the data versus 40 time from an acceleration sensor or sensors. 41 42 Secondary containment. The structure surrounding the primary containment that 43 acts as a further barrier to control the release of radioactive material. 44 8

l Seismic Trigger. A device that starts the time-history accelerograph. 2 3 Time-History Accelerograph. An instrument capable of measuring and 4 permanently recording the absolute acceleration versus time. The components 5 of the time-history accelerograph (acceleration sensor, recorder, seismic 6 trigger) may be assembled in a self-contained unit or may be separately 7 located. 8 9 Triaxial. Describes the function of an instrument or group of instruments in 10 three mutually orthogonal directions, one of which is vertical. 11 12 13 16 17 9

1 REGULATORY ANALYSIS 2 3 A separate regulatory analysis was not prepared for this regulatory 4 guide. The draft regulatory analysis, "Proposed Revision of 10 CFR Part 100 5 and 10 CFR Part 50," provides the regulatory basis for this guide and examines 6 the costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, 9 as Enclosure 2 to secy 92-215. 10 10

Suggested Changes Either Highlighted or Bold. October 1992 Division l Draft DG-1017

Contact:

R. M. Kenneally (301) 492-3893 DRAFT REGULATORY GUIDE DG-1017 PRE-EARTHQUAKE PLANNING AND IMMEDIATE NUCLEAR POWER PLANT OPERATOR POST#.EARTHQUAKE

                                             *:*:~

ACTIONS -2 l A. INTRODUCTION 3 Paragraph IV(a)(4) of Proposed Appendix s, "Earthquake Engineering 4 Criteria for Nuclear Power Plants," to 10 CFR Part SO, "Domestic Licensing 5 of Production and Utilization Facilities," would require that suitable 6 instrumentation' be provided so that the seismic response of nuclear power 7 plant features important to safety can be evaluated promptly. Paragraph 8 IV(a)(3) of Proposed Appendix S to 10 CFR Part 50 would require shutdown of 9 the nuclear power plant if vibratory ground motion exceeding that of the 10 operating basis earthquake ground motion (OBE) or significant plant damage 11 occurs. Proposed Paragraph S0.54(ee) to 10 CFR Part 50 would require 12 licensees of nuclear power plants that have adopted the earthquake 13 engineering criteria in Proposed Appendix s to 10 CFR Part 50 to shut down 14 the plant if the criteria in Paragraph IV(a)(3) of Proposed Appendix s are 17 exceeded. This guide is being developed to provide guidance acceptable to the NRC staff for a timely evaluation after an earthquake of the recorded 18 instrumentation data and for determining whether plant shutdown would be 19 required by the proposed amendments to 10 CFR Part so. 20 Any information collection activities mentioned in this draft regulatory 21 guide are contained as requirements in the proposed amendments to 10 CFR Part 22 50 that would provide the regulatory basis for this guide. The proposed 23 amendments have been submitted to the Office of Management and Budget for 24 clearance that may be appropriate under the Paperwork Reduction Act. Such 25 clearance, if obtained, would also apply to any information collection 26 activities mentioned in this guide. 1 27 Guidance is being developed in Draft Regulatory Guide DG-1016, the Second 28 Proposed Revision 2 to Regulatory Guide 1.12, "Nuclear Power Plant Instru-29 mentation for Earthquakes," to describe seismic instrumentation acceptable to 30 the NRC staff. 31 32 1

1 B. DISCUSSION 2 3 When an earthquake occurs, ground motion data are recorded by the seismic 4 instrumentation.' These data are used to make an early determination of the 5 degree of severity of the seismic event. The data from the seismic 6 instrumentation, coupled with information obtained from a plant walkdown, are 7 used to make the initial determination of whether the plant should be shut 8 down, if it has not already been shut down by operational perturbations 9 resulting from the seismic event. If on the basis of these initial evalua-10 tions (instrumentation data and walkdown) it is concluded that the plant shut-11 down criteria have not been exceeded, it is presumed that the plant will not 12 be shut down. Guidance is being developed on postshutdown inspections and 13 plant restart; see Draft Regulatory Guide DG-1018, "Restart of a Nuclear Power a: 16 Plant Shut Down by a Seismic Event." The Electric Power Research Institute has developed guidelines that will enable licensees to quickly identify and assess earthquake effects on nuclear 17 power plants. These guidelines are in EPRI NP-5930, "A Criterion for Deter-18 mining Exceedance of the Operating Basis Earthquake," July 1988; EPRI NP-6695, 19 "Guidelines for Nuclear Plant Response to an Earthquake," December 1989 2 ; and 20 EPRI TR-100082, "Standardization of Cumulative Absolute Velocity," December 21 1991. 2 22 This regulatory guide is based on the assumption that the nuclear power 23 plant has operable seismic instrumentation, including the equipment and soft-24 ware required to process the data within 4 hours after an earthquake. This is 25 necessary because the decision to shut down the plant will be made, in part, 26 by comparing the recorded data against OBE exceedance criteria. The decision to shut down the plant is also based on the results of the pf.jfi§ walkdown inspections that take place within 8 hours of the event. If the seismic 29 instrumentation is inoperable, the guidelines in Appendix A to this guide 30 would be used to determine whether the operating basis earthquake ground 31 motion (OBE) has been exceeded. 32 Shutdown of the nuclear power plant may be required if the vibratory 33 ground motion experienced exceeds that of the OBE . Two criteria for determin-34 ing exceedance of the OBE are provided in EPRI NP-5930: a threshold response 35 spectrum ordinate criterion and a cumulative absolute velocity criterion 36 (CAV). A procedure to standardize the calculation of the CAV is provided in 37 EPRI TR-100082. A spectral velocity threshold has also been recommended by 38 EPRI since some structures have fundamental frequencies below the range speci-39 fied in EPRI NP-5930. The NRC staff now recommends 1.0 to 2.0 Hz for the 40 range of the spectral velocity limit since some structures have fundamental 41 frequencies below 1.5 Hz. The former range was 1.5 to 2.0 Hz. 2 42 EPRI reports may be obtained from the Electric Power Research Institute, 43 Research Reports Center, P.O. Box 50490, Palo Alto, CA 94303 2

1 The NRC staff does not endorse the philosophy discussed in EPRI NP-6695, 2 Section 4.3.4 (first paragraph, last sentence), pertaining to plant shutdown 3 considerations following an earthquake based on the need for continued power 4 generation in the region. If the licensee determines that plant shutdown is 5 required by the Commission's regulations, but the licensee does not consider 6 it prudent to do so, the licensee may ask for an emergency exemption from the 7 requirements of the regulation pursuant to 10 CFR Part 50.12 so that the plant 8 need not shut down if the exemption is granted. 9 Appendix B to this guide provides definitions to be used with this 10 guidance. 11 12 C. REGULATORY POSITION 13

1. BASE-LINE DATA 16 1.1 Information Related to Seismic Instrumentation 17 18 A file containing information on all the seismic instrumentation should 19 be kept at the plant. The file should include:

20 21 1. Information on each instrument type such as make, model, and serial 22 number; manufacturers' data sheet; list of special features or options; per-23 formance characteristics; examples of typical instrumentation readings and 24 interpretations; operations and maintenance manuals; repair procedures (manu-25 facturers' recommendations for repairing common problems); and a list of any 26 special requirements, e.g., maintenance, operational, installation. 27 ~ 2. Plan views and vertical sections showing the location of each 29 seismic instrument and the orientation of the instrument axis with respect to 30 a plant reference axis. 31 32 3. A complete service history of each seismic instrument. The service 33 history should include information such as dates of servicing, description of 34 completed work, and calibration records and data (where applicable). 35 36 4. 37 cumulative absolute velocity (see Regulatory Position 4). These data should 38 be obtained after the initial installation and each servicing of the 39 containaent foundation instrumentation using a suitable earthquake time-40 history (e.g., the October 1987 Whittier, California, earthquake) or 41 manufacture's calibration standard. 42 43 1.2 Planning for Postearthguake Inspections 44 3

1 The selection of equipment and structures for inspections and the content 2 of the baseline inspections as described in Sections 5.3.1 and 5.3.2.1 of EPRI 3 NP-6695, "Guidelines for Nuclear Plant Response to an Earthquake," are accept-4 able to the NRC staff for satisfying the proposed requirements in Paragraph s IV(a)(3) of Proposed Appendix s to 10 CFR Part 50 for ensuring the safety of 6 nuclear power plants. 7 8 2. IMMEDIATE POSTEARTHOUAKE ACTIONS 9 10 The guidelines for immediate postearthquake actions specified in Sections 11 4.3.l and 4.3.2 (including Section 5.3.2.1 and items 7 and 8 of Table 5-1) of 12 EPRI NP-6695 are acceptable to the NRC staff for satisfying the requirements 13 proposed in Paragraph IV(a)(3) of Proposed Appendix s to 10 CFR Part SO. 14

• 15 16 17 18 19 3.

3.1 EVALUATION OF GROUND MOTION RECORDS Data Identification A record collection log should be maintained at the plant, and all data 20 should be identifiable and traceable with respect to: 21 22 1. The date and time of collection, 23 24 2. The make, model, serial number, location, and orientation of the 25 instrument (sensor) from which the record was collected. 26 27 3.2 Data Collection 28 - 29 3.2.l Only personnel trained in the operation of the instrument should 30 collect the data. 31 32 3.2.2 The steps for removing and storing records from each seismic 33 instrument should be planned and performed in accordance with established 34 procedures. 35 36 3.2.3 Extreme caution should be exercised to prevent accidental damage 37 to the recording media and instruments during data collection and subsequent 38 handling. 39 40 3.2.4 As data are collected and the instrumentation is inspected, notes 41 should be made regarding the condition of the instrument and its installation, 42 for example, instrument flooded, mounting surface tilted, fallen objects that 43 struck the instrument or the instrument mounting surface. 44 4

1 3.2.S For validation of the collected data, a reference signal (see 2 Regulatory Position 1.1(4)) should be added to the record without affecting 3 the previously recorded data. 4 5 3.2.6 If the instrument ' s operation appears t o have been normal , the 6 instrument should remain in service wi thout read j ustment or change that would 7 defeat attempts to obtain postevent calibration. 8 9 3.3 Record Evaluation 10 11 Records should be analyzed according to the manufacturer ' s specifications 12 and the results of the analysis should be evaluated . Any record anomalies , 13 invalid data, and nonpertinent signals should be noted, along with any known causes. 16 4. DETERMINING OBE EXCEEDANCE 17 18 The evaluation to determine whether the OBE was exceeded should be 19 performed using data obtained from the three components of the containment 20 foundation tri-axial accelerometer (i. e ., two horizontal and one vertical). 21 The evaluation may be performed on uncorrected earthquake records. It was 22 found in a study of uncorrected versus corrected earthquake records (see EPRI 23 NP-5930) that the use of uncorrected records i s conservative. The evaluation 24 should consist of a check of the response spectrum, cumulat ive absolute 25 velocity limit, and the operability of the instrumentation. IJHJ:II)i:f:jjf!fjp;§fi 26 **ai,ww,.q@J?\ttF!=Jf.tl}tliTtnt=im+no.u.hs.*=====#t,t:tt~*******~*=#Affl!ik~n*'Qc.P.ume.e.;.(i

4. 1 Response Spectrum Check 29 30 The OBE response spectrum is exceeded if any one of the three components 31 (two horizontal and one vertical) of the 5 percent damped containment 32 foundation response spectra §iJ.iJ,Jii.ii~U(:iiHilJ.fi=i=i.f:@f=#iQ.'i!si.i.iKi.PP.ifi&#M.iii 33 *t.-1alllaif.Sle.tJmlJ!.@=l*~t~*¥.s@mgf.,!!:@!'#:m:t~@!P:t:$t~l:=i.Pfa§g; is larger than:

34 35 1. The corresponding design response spectral acceleration (OBE 36 spectrum if used, otherwise 1/3 of the safe shutdown earthquake 37 (SSE) spectrum) or 0.2g, whichever is greater, for frequencies 38 between 2 to 10 Hz, or 39 40 2. The corresponding design response spectr al velocity ( OBE spectrum if 41 used, otherwise 1/3 of the SSE spectrum) or a spectral velocit y of 6 42 inches per second, whichever is greater, for frequencies between 1 43 to 2 Hz. 44 5

1 4.2 Cumulative Absolute Velocity (CAV) Limit 2 3 For each component of the containment foundation sensor the CAV should be 4 calculated as follows: (1) the absolute acceleration (g units) time-history 5 is divided into 1-second intervals, (2) each 1-second interval that has at 6 least 1 exceedance of 0.025g is integrated over time, (3) all the integrated 7 values are summed together to arrive at the CAV. The CAV limit is exceeded if 8 any CAV calculation is greater than 0.16 g-second. Additional information on 9 how to determine the CAV is provided in EPRI TR-100082. 10 11 4.3 Instrument Operability Check 12 13 After an earthquake at the plant site, the response spectrum and CAV should be obtained using the calibration standard (see Regulatory Position 1.1 ( 4)) to demonstrate that the liMtMi~l#.#,Xi@f.P.1.iJ.i1?:t.-.m!(glJ~ii#*it~Hl~@J:iJ.#f.W&! 16 was functioning properly. 17 18 4.4 Inoperable Instrumentation 19 20 If the seismic instrumentation is inoperable, the criteria in Appendix A 21 to this guide should be used to determine whether the OBE has been exceeded. 22 23 5. CRITERIA FOR PLANT SHUTDOWN 24 25 If the OBE is exceeded or significant plant damage occurs, plant shut 26 down ma Y be required * :fatW:@s#.:i.mt:MWPldt:§);if.J{g~l!ig;,::tU~IP.~i!$f.,::,,.,%gWP.#.f.iPP!W#.9#P.J1!~41 ea 27 29 mi}).; ~% 5.1 OBE Exceedance 30 31 If the response spectrum check and the CAV limit (performed in accordance 32 with Regulatory Positioni. 4.1 and 4.2) were exceeded, e th.in

                                                                                       *****~*x***

the OBE was 33 exceeded and plant shutdown may be required. If either limit does not exceed 34 the criterion, the earthquake motion did not exceed the QBE. llilen:!l@t!bi 35 @*Me,w.:1-..,,.,,.a~F,t:tt~i\PJVtEJm;.:tt==c.~f<1!i=tc.hecxii:a:ft*f!t1b;:a,-i::&fam44@M, 36 @.@1§.i.lf:ki{!JA.i@jllf.~WM!'i~g;tijb.1Ul~¢.-~~nP.J:t,P.#$t)~#,~!jj'fP.f:,:::ft.fi':J.,,-,:,:,,p:..-.+s#k9~4.fM The 37 determination of whether or not the OBE has been exceeded should be performed 38 even if the plant automatically trips off-line as a result of the earthquake. 39 5.2 Damage 40 41 The plant should be shut down if the walkdown inspections, performed in 42 accordance with Regulatory Position 2, discover damage ,,r==l.~#:t~Hkiin,:,:,,,,trnot 43 ~nt,-:$llMMiift~Pi@ilMJM~

• 44 6

1 6. PRE-SHUTDOWN INSPECTIONS 2 3 The pre-shutdown inspections described in Section 4.3.4 of EPRI NP-6695, 4 "Guidelines for Nuclear Plant Response to an Earthquake," ' with the last sen-s tence in the first paragraph of section 4.3.4 deleted, are acceptable to the 6 NRC staff for satisfying the requirements proposed in Paragraph IV(a)(3) of 7 Proposed Appendix s to 10 CFR Part 50 for ensuring the safety of nuclear power 8 plants. 9 The following paragraph in Section 4.3.4 of EPRI NP-6695 is repeated to 10 emphasize that the plant should shut down in an orderly manner. 11 12 "Prior to initiating plant shutdown following an earthquake, visual 13 inspections and control board checks of safe shutdown systems should be performed by plant operations personnel, and the availability of off-site and emergency power sources should be determined. The pur-16 pose of these inspections is to determine the effect of the earth-17 quake on essential safe shutdown equipment which is not normally in 18 use during power operation so that any resets or repairs required as 19 a result of the earthquake can be performed, or alternate equipment 20 can be readied, prior to initiating shutdown activities. In order 21 to ascertain possible fuel and reactor internal damage, the follow-22 ing checks should be made, if possible, before plant shutdown is 23 initiated * * * * " 24 25 If the OBE was not exceeded and the walkdown inspection P.Qh#NG,J,j4 26 @llliK~ll':J(J!:tf#WsDr.=id;ffiaiiifi!#swrrejc.~Jas1hie!CJ!jJ\pedt#Ph:~lljtf~l:#!~'1~ es 27 29 indicates no damage to the-nuclear power plant, shutdown of the plant is not required. The plant may continue to operate (or restart following a post-trip review, if it tripped off-line because of the earthquake). 7

1 D. IMPLEMENTATION 2 3 The purpose of this section is to provide guidance to applicants and 4 licensees regarding the NRC staff ' s plans for using this regulatory guide. 5 This draft guide has been released to encourage public participation in 6 its development. Except in those cases in which the applicant proposes an 7 acceptable alternative method for complying with the specified portions of the 8 Commission's regulations, the method to be described in the active guide 9 reflecting public comments will be used in the evaluation of applications for 10 construction permits, operating licenses, combined licenses, or design certi-11 fication submitted after the implementation date to be specified in the active 12 guide. This guide would not be used in the evaluation of an application for 13 an operating license submitted after the implementation date to be specified -14 in the active guide if the construction permit was issued prior to that date. 8

1 APPENDIX A 2 INTERIM OPERATING BASIS EARTHQUAKE EXCEEDANCE GUIDELINES 3 4 Thia regulatory guide is based on the assumption that the nuclear power 5 plant has operable seismic instrumentation. If the seismic instrumentation is 6 inoperable, the following should be used to determine whether the operating 7 basis earthquake ground motion (OBE) has been exceeded: 8 9 1. Deteraination of OBE exceedance is based on the response spectrua check 10 described in Regulatory Position 4.1 of this regulatory guide. A com-11 parison is aade between the foundation-level design response spectra and 12 data obtained from the foundation-level instruments. If the response 13 spectrua check at any foundation is exceeded, the OBE is exceeded and !fij ftlll!If!lilIiiii!IB

• 16 2. For plants at which no instrumental data are available, the OBE will be 17 considered to have been exceeded and #.ff.lf,¥.iffltHillYf~f)J.lffiffif~ if one of 18 the following applies:

19 20 1. The earthquake resulted +.ttfi#:fl!ll?ilim!iit:l:!!!l!!lil{Hitl.iiiB!#I:!i~ 21 ttf:l#.iilltll.1¥111:$.J~ 22 23 2. The earthquake was lmli:::::!l!!linlll!:le!l!f?Ii:il:Ht!! of magnitude 6. 0 or 24 greater, or 25 26 3. The earthquake was felt within the plant and was of magnitude 5.0 or greater and occurred within 200 km of the plant. 29 3. A postearthquake plant walkdown should be conducted (see Regulatory 30 Position 2 of this guide). 31 32 4. If plant shutdown is warranted under the above guidelines, the plant 33 should be shut down in an orderly manner (see Regulatory Position 6 of 34 this guide)

• 35 36 Note:

37 The determinations of epicentral location Md magnitude by the u.s. 38 Geological Survey, National Earthquake Information Center, will usually 39 take precedence over other estimates; however, regional and local 40 determinations will be used if they are considered to be more accurate. 41 Also, higher quality damage reports or a lack of damage reports from the 42 nuclear power plant site or its immediate vicinity will take precedence 43 over more distant reports. 44 A-1

1 APPENDIX B 2 DEFINITIONS 3 4 Design Response Spectra. Response spectra used to design Seismic Category I 5 structures, systems, and components. 6 7 Operating Basis Earthcµiake Ground Motion (OBE). The vibratory ground motion a ,,-w*1,,D-K~J,@w,,e1AiliJ~tAt.ia+ffi*i,.g¢.;;$@tn~#lt~~,,m:,*~¥~+9.a¥t.Y::,:,::*-..J.iiP.#.,ta+t>Y 9 S:ffWt;lW!,f!!rt~@;f!@J!Ili!:f:ie:lf1/4fiifHE

• 10 11 Spectral Acceleration. The acceleration response of a linear oscillator with 12 prescribed frequency and damping.

13 Spectral Velocity. The velocity response of a linear oscillator with pre-scribed frequency and damping. 16 17 18 19 20 21 22 23 24 25 26 B-1

1 REGULATORY ANALYSIS 2 3 A separate regulatory analysis was not prepared for this regulatory 4 guide. The draft regulatory analysis, "Proposed Revisions of 10 CFR Part 100 5 and 10 CFR Part 50," provides the regulatory basis for this guide and examines 6 the costs and benefits of the rule as implemented by the guide. A copy of the 7 draft regulatory analysis is available for inspection and copying for a fee at 8 the NRC Public Document Room, 2120 L Street NW. (Lower Level), Washington, DC, 9 as Enclosure 2 to Secy 92-215. 10 11 12 13 16 17 18 19 20 21 22 23 24 25 26 29 30 RA-1

Pl r o 5 1. J-1 o DOCKET NUMBER p~pOSED RULE C51 Ffl '17 'irtJ 2) 1 0

                                                                                  '.'t.( _1.5 OOCICETEo MAit 2 4 1993 March 22, 1993                                                              ~I SEJIIYICE lfWQ 9£CY~

COMMENTS OF OHIO CITIZENS FOR RESPONSIBLE ENERGY, INC. - - ON PROPOSED RULE, "REACTOR SITE CRITERIA; I NCLUDlN.....-. * .w..

                                                                                  * ...,_..,.. ... v EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS AND PROPOSED DENIAL OF PETITION FOR RULEMAKING FROM FREE AMERICA, INC. ET AL.", 57 FED. REG. 47802 (OCTOBER 20, 1992)

OCRE generally supports the revisions to the regulations proposed herein (with some exceptions), and proposes some enhancements which would improve safety. To the extent applicable, these comments should be considered as addressing the set of draft Regulatory\ Guides and Standard Review Plan section related to this rulemaking as well (DG-1015, DG-1016, DG-1017, DG-1018, DG-4003, and proposed Revision 3 to SRP 2.5.2). I. Reactor Site Criteria (Nonseismic) A. Background and General Approach OCRE supports the decoupling of siting suitability decisions from plant design. 10 CFR Part 100, when promulgated in 1962, may have intended to preclude metropolitan siting of nuclear power reactors; however, in practice, the regulation has had a greater influence on plant design than on actual site population charac-teristics. Part 100 does not include any demographic criteria or fixed distances for siting. Rather, site suitability is deter-mined by calculating doses to hypothetical individuals stationed at the Exclusion Area Boundary for 2 hours and at the Low Popula-tion Zone Boundary for the duration of the accident (taken in regulatory guidance to be 30 days). The size of these boundaries (and consequently the Population Center Distance) are determined by a combination of engineered design features which limit radio-nuclide release and site meteorology. Dr. David Okrent, a longtime former member of the ACRS, has thoroughly documented the difficulties of applying Part 100 in a manner which excludes populated areas. In his book Nuclear Reactor Safety: Qn :t.ha History o.f. :t.ha Regulatory Process (1981, University of Wisconsin Press), Dr. Okrent states, "it appears that the flexibility of the site criteria permitted an encroach-ment by utilities on the stated AEC policy of keeping power reactors away from densely populated centers. This was done by substituting engineered safeguards for distance within the con-text of an independent, last-ditch protection afforded by the 1 MAY 111993 ~ Acknowledged by card ................................"

U.S. NUCLt.**,r-, ~i.(1ULATORY COMMISSIO~ OOCKETING & SEPWICE SECT~N OFFICE OF THE SECRETAIW Of THE COMMISSION

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DIIUmll"lt Statistics Postmark Date Jh-7-1&:tJ F ~xd ~ 3 /~.,_JqJ Copies Received I Add'ICopiisRepr_od_uc_ad___;..;

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Sp~eial DistritxJMon fl.T: rJ S , f.) (}l?p_

• _M_ 1,,4.,11p l,.. YJ -r~~J o a ~

containment building." Id. at 46-47. Clearly the most notorious attempt at metropolitan siting was the proposed Ravenswo9d reac-tor, which was to have been sited in the Borough of Queens, 'New York City. The utility justified the site by reliance on engi-neered safeguards, principally a "zero-leakage" double contain-ment. It was not until 1966 that an understanding began to develop in the reactor safety community that a "major accident resulting in substantial meltdown of the core" (the accident for which the fission product release of Part 100 is postulated) would result in forces which could challenge containment integrity. "It seems fair to assume that in the spring of 1965, if there were groups or individuals who made a direct connection between large-scale core melting and a loss of containment integrity, it was not a widely held piece of knowledge." Okrent, p. 110. By June 1966 the AEC regulatory staff "recognized the inconsistency of the staff's position that, on one hand, the core would melt, leading to fission product release and metal-water reaction considera-tions, but that, on the other, core melt would not cause a prob-lem from a loss of containment integrity." Id. at 115. Thus, the engineered safeguards (principally containment) that were the justification for the acceptance of poor sites would likely fail to maintain the low leak rates proposed (on the order of

0. 1%/day) in precisely the accident situations where they were most needed.

Experiences such as Ravenswood, coupled with the evolving knowl-edge of reactor accidents, led the staff to adopt regulatory guidance which placed limits on actual population density for reactor sites, e.g., Regulatory Guide 4.7. However, regulatory guides are not regulations and compliance with them is not manda-tory; rather, they represent a method acceptable to the staff for compliance with the regulations. GYli.. States Utilities (River Bend Station, Units 1 and 2), ALAB-444, 6 NRC 760 (1977). If the NRC wants to prohibit siting of nuclear power plants in densely populated areas, it should do so explicitly in an enforceable regulation. Therefore, OCRE supports decoupling siting from design and the specification of demographic criteria for siting. OCRE also supports the elimination of the requirement for the LPZ as being superseded by the NRC's emergency planning regulations which require a 10-mile EPZ. The Population Center Distance likewise should be replaced with population density criteria. B. Population Density Criteria The question thus before us is what are the appropriate numerical 2

criteria for EAB size and population density. A number of fig-urea have been proposed, in this proposed rule and elsewhere. Most appear to be arbitrary. For example, in 1950 the ACRS proposed a rule of thumb for calculating the EAB radius (Okrent,

p. 18):

R = 0.01 x SQRT(P) where R is the EAB radius in miles and Pis the power level in kWt. For a 3800 MWt reactor, according to this formula, the RAB radius should be about 19.5 miles. The idea was that beyond radius R, the calculated radiation exposure should be less than 300 rem (whole body}, or that evacuation should be possible. Of course, this formula assumed that the reactor had no containment. It was the large exclusion distances calculated by this rule of thumb which prompted the concept of containment so that those distances could shrink.. In May 1967, after knowledge of the "China syndrome" had removed containment as an independent bulwark, the ACRS proposed the following population density limit (Okrent, p. 140): Ptot(R) ~ 4000 x RA2 for current population Ptot(R) ~ 5000 x RA2 for population projected 25 years hence where 5 ~ R 25 in miles and Ptot(R) is the total population within a distance R miles from the reactor. The purpose of this formula was to place a moratorium on sites worse than Indian Point. Dividing these formulas by the area of a circle with radius R, Pi x RA2, to get the equivalent population densities yields: 4000/Pi = 1273 persons/sq. mi. for current popula-tion, and 5000/Pi = 1592 persons/sq. mi. for population projected 25 years hence. These numbers are well above those contained in the proposed rule. The 1979 Siting Policy Task Force Report, NOREG-0625, recommended the following values for population density: From the EAB to 5 miles, the population density at the beginning of reactor operation should not exceed one-half of the average 3

population density ,of the region, or 100 persons per square mile, whichever is greater. The population within this annulus should not be expected to double during the life of the plant, and no more than half of the allowed number of persons in the zone should be permitted within any single 22-1/2 degree sector. Transients should be weighted according to their fractional occupancy within this annulus. From 5 to 10 miles, the population density at the beginning of reactor operation should not exceed 3/4 of the average population density of the region, or 150 persons per square mile, whichever is greater. From 10 to 20 miles, the population density at the beginning of reactor operation should not exceed twice the average population density of the region, or 400 persons per square mile, whichever is greater. For both the 5-10 mile and 10-20 mile annular rings, the Task Force repeated the restriction on concentration of the population in any particular 22-1/2 degree sector that was articulated in the proposal for the inner ring, but did not include restrictions for projected growth or weighting of transients. The Task Force did not see the need for limits on population beyond 20 miles from the site. The proposed rule would codify the Regulatory Guide 4.7 stand-ards: EAB of at least 0.4 miles, and initial population density limit of 500 persons/sq. mi. averaged over any radial ~istance out to 30 miles, with a projected population density limit (40 years hence) of 1000 persons/sq. mi. averaged over any radial distan0e out to 30 miles. Weighted transient populations must be included in these calculations. In OCRE's view, the most problematic aspect of setting demograph-ic criteria lies in the projection of future population growth up to 40 years into the future. The NRC correctly notes that the validity and reliability of such projections decrease markedly as the projection time increases. However, it is OCRE's position that the entire plant lifetime must be considered when making these projections, since the NRC has adopted license renewal regulations, 10 CFR Part 54. It is especially important to consider the period of license renewal at the outset of licensing because of the doctrine set forth in Part 54: the adequacy of the current licensing basis. The fact is that site population char-4

acteristics will not be revisited in the license renewal review, so that period of operation must be considered at the outset. While it is certainly true that population projections beyond 40 years into the future are speculative, the same can be said for such projections even 20-30 years from now. It is important to consider the fact that a nuclear plant may have a "positive feedback" effect on local population growth. In addition to the influx of plant employees into the area, the plant may offer various benefits to the community which attract people. For example, consider the Perry Nuclear Power Plant in Lake County, Ohio. Due to the plant, residents in the Village of North Perry get free cable TV, free water, and free garbage collection. The Perry school district has received copious tax benefits, which has resulted in a superior quality school system, which attracts families with school age children. Presently that school system is building a new high school which resembles a small community college campus, complete with Olympic-size swim-ming pools. Consequently, residential development in the Perry area is increasing. OCRE believes that the best way to ensure that population levels do not increase beyond acceptable limits throughout the life of the plant is to require, at the outset, sufficiently stringent population density criteria and remote siting so that it is unlikely that population growth will reach unacceptable levels throughout the plant life. OCRE proposes that the NRC adopt the specific numerical criteria, modeled on the 1979 Siting Policy Task Force, modified to read as follows: From the EAB to 5 miles, the population density at the beginning of reactor operation should not exceed 100 persons per square mile. From 5 to 20 miles, the population density at the beginning of reactor operation should not exceed 150 persons per square mile. From 20 to 40 miles, the population density at the beginning of reactor operation should not exceed 300 persons per square mile. The population within this region should not be expected to double during the life of the plant, and no more than half of the allowed number of persons in each annular zone should be permit-ted within any single 22-1/2 degree sector. Transients should be weighted according to their fractional occupancy ~ithin each annulus. 5

OCRE's reasons for extending the Task Force's recommended 20 miles to 40 miles are twofold: first, to minimize early fatali-ties and second, to ensure a sufficient distance from urban areas. According to NUREG/CR-2239, population centers beyond 25 miles do not contribute to early fatalities (p. 2-89). However, population centers between 10 and 20 miles can contribute to early fatalities due to rainout of the plume; hence, OCRE's extension of the 5-10 mile annulus out to 20 miles. Adopting a 40 mile distance for demographic criteria, as well as an initial limit of 300 persons per square mile, makes it less likely that the site will be engulfed by suburban sprawl, and the longer commuting distance makes the site area less attractive as a bedroom community. The Federal Register notice states that population density re-strictions out to 40 miles could make it difficult to obtain suitable sites in some areas of the country. OCRE believes that this should not be a concern of the NRC. The NRC's sole mandate is to ensure safety, not the availability of nuclear reactor sites. These numerical criteria should be specified in the regulation itself, not in unenforceable regulatory guidance. These criteria should be considered strict upper limits of acceptability. C. Exclusion Area Boundary For the minimum EAB radius, OCRE would propose a distance of 1.0 mile. The basis for this distance is twofold: first, to mini-mize early fatalities, and second, to expand the zone of control by the licensee to exclude potential terrorist attackers. NUREG/CR-2239 notes that, for source term SST1 reduced tenfold, on the average fatalities would be confined to 1 mile. For the SST2 source term, early fatalities would be confined to 0.5 miles. It is concluded that for releases substantially smaller than SST1, a 1 mile EAB can have a substantial impact even with-out an emergency response. NUREG-0625 also noted that increasing the EAB to one mile would "provide significant additional protec-tion against Class 9 accidents" (p. 47). OCRE believes that the EAB should serve not only to protect the public from the reactor, but also to protect the reactor from malevolent persons in society. A minimum EAB radius of 1.0 mile, within which the licensee has total control of all activities through ownership of property and the application of appropriate security measures, could help minimize the threat of terrorist acts of radiological sabotage. Two recent events have illustrat-ed the vulnerability of nuclear power plants to terrorist attack: 6

(1) the February 7, 1993 event at Three Mile Island where an individual drove a vehicle onto the site and crashed a gate into the protected area (the individual was not apprehended for 4 hours); and (2) the February 26, 1993 terrorist bombing of the World Trade Center in New York. While the latter event was not directed against a nuclear facility, it certainly illustrates that terrorists can operate effectively within the United States, can gain access to explosives, and can cause destruction and death. The event at TMI raises some serious "what if" questions: what if the individual had been driving a vehicle laden with explosives? What if the individual, who was loose on the site for 4 hours, had been armed? What if the individual had been accompanied by a team of armed commandos in the vehicle? To provide appropriate protection, the EAB should not be tra-versed by any highways, railroads, or waterways on which traffic is freely permitted. This may present special problems on water-ways. Most nuclear power plants are located adjacent to naviga-ble waterways. Prohibiting boat traffic within 1 mile of the nuclear plant may require legislative authority. Under the civil and common law, free public use of navigable waters takes prece-dence over riparian or littoral rights. (See 78 Am Jur 2d Sec-tions 86-112.) OCRE believes that restrictions on public use of waterways within 1 mile of a nuclear power plant is necessary to preclude terrorist use of watercraft to approach the plant. D. Sites with Multiple Reactors Proposed 10 CFR 100.21(a) (2) states: "If the reactors are inter-connected to the extent that an accident in one reactor would initiate an accident in another, the size of the exclusion area for each reactor must be determined on a case by case basis." (Emphasis added. ) Use of the word "would" means that such acci-dent *initiation must be a certainty. It would be more appropri-ate to incorporate the language of the current Part 100. 11(b) (2). OCRE proposes that the sentence quoted above be revised to read: "If the reactors are interconnected to the extent that an acci-dent in one reactor could affect the safety of operation of any other, the size of the exclusion area must be determined on a case by case basis." This revision is necessary because some of the modular reactor designs being proposed have multiple reactors being controlled from a common control room with a very small crew of operators. It is not hard to envision how, in such a configuration, an accident in one reactor could adversely affect the safety of the others simply by diverting operator attention from the other reactors. (For a discussion of the safety implications of the 7

use of a common control room and small operator crews for these modular designs, MHTGR and PRISM, see Advanced Reactor Study, prepared for the Onion of Concerned Scientists by MHB Technical Associates, July 1990, Section 3.4.) E. Community Consent While this may not be within the NRC's statutory jurisdiction, OCRE believes that community consent ought to be a prerequisite for siting a nuclear power plant. This consent would consist of a an affirmative vote on a ballot issue by the ma.Jority of the voters in all local government jurisdictions within 10 miles of the proposed reactor. OCRE believes that simple fai~ness and the principles of a democratic society demand nothing less. F. Consideration of Natural Phenomena and Hazards The NRC should consider, at the siting stage, all possible natu-ral phenomena at the site and their effects on plant safety. One such phenomenon which appears to have been overlooked is forest fires. The NRC should also consider, at the siting stage, the potential impacts of all possible natural phenomena and man-related hazards on emergency planning. For example, the lessons learned from the experience of Hurricane Andrew at Turkey Point should be incorpo-rated into the siting process. The impact of earthquakes on emergency planning must also be considered. OCRE is aware that the NRC has rejected an interpretation of its emergency planning regulations which would have considered this impact. However, OCRE believes that it is a different question to consider such potential impediments to evacuation at the siting stage, rather than the question that was posed at the operating license stage regarding implementation of emergency planning requirements. G. Siting Policy Task Force Recommendations OCRE herein addresses those recommendations of the Siting Policy Task Force (which have not been considered previously) which deserve further evaluation by the NRC. Recommendations 2 and 6 These recommendations are related in that they have as their objective the selection of reactor sites with favorable charac-teristics, rather than the reliance on reactor designs to compen-sate for less desirable sites. "The Task Force believes that 8

there is merit to maintaining the safety factor inherent in physical distance and that the distance £actor should not be traded off £or design features of the plant." NUREG-0625, p. 51. Since one of the purposes of the proposed rule is to decouple siting from plant design, it would behoove the NRC to reconsider its disposition of these recommendations. The Task Force's concern that applicants should be encouraged to select more favorable sites has validity. Recommendation 5 The NRC is requesting comment on one of the subparts of this recommendation: that licensees be required to monitor and report potentially adverse offsite developments. OCRE believes this should be required. Such offsite developments should include new nearby (within 5 miles) potentially hazardous industrial facilities and changes in population density in the area. While NUREG-0625 cited a 1977 NRC inspection procedure for reviewing changes to plant environs on a 3-year cycle (p. 21), it is not apparent that it is still in effect, nor is it obvious to OCRE that this receives a very high priority, or that it has actually been implemented at some plants, e.g., Perry. Requiring the licensee to monitor and report changes in offsite land use will alert the NRC to potentially important changes when they occur, so that they can get the regulatory attention needed. Possible regulatory actions which could be taken in response to such changes in offsite land use include reductions in power level, enhanced engineered safeguards, or other compensatory measures. The NRC's continuing attention to local offsite developments is especially important given the doctrine set forth in the NRC's license renewal rule, 10 CFR 54, on the adequacy of the current licensing basis. This rule assumes that the NRC continues to diligently monitor such developments after a nuclear plant re-ceives an operating license. The NRC should validate this as-sumption by adopting these proposed requirements. One of the ideas advanced in NUREG-0625, that legislation be enacted which would give the NRC control over local land use, ought to be studied further. This option is one mechanism that would ensure that hazardous industrial facilities are not de-veloped near nuclear power plants in the post-licensing stage. Recommendation 8 9

This recommendation proposed that a final decision disapproving a proposed site by a state agency whose approval is fundamental to the project would be a sufficient basis for the NRC to terminate review. The termination of a review would then be reviewed by the Commission. The reason given in the Federal Register notice for the rejection of this recommendation is that it "would give a State the author-ity to grant issuance of a construction permit for a nuclear facility. Only the Federal Government has this authority." This "logic" is exceedingly flawed. The right of a State to deny approval of a site is not the same as the ability to grant a construction permit. NUREG-0625 advances valid reasons for this recommendation: "The Task Force believes that there is little useful purpose to be served by NRC's continuing to review a project if required State approval of the proposed site has been denied. Provisions in the siting regulations on this matter would be beneficial because (a) resources could be applied to viable alternatives and not wasted on a fait accompli, and (b) the regulations would recognize and enhance the role of State governments in the site selection and approval process" (p. 61). The Task Force imposed safeguards on applicants' rights by including the provision for Commission review. This recommendation should be adopted. II. Seismic and Earthquake Engineering Criteria A. General Comments OCRE supports the revisions relating to the OBE and the require-ment for plant shutdown if the OBE is exceeded as appropriate and justified. OCRE also supports the administrative changes, i.e., relocating earthquake engineering criteria from Part 100 to Part 50. OCRE also supports the revisions to the terminology employed in Part 100 Appendix B to be more applicable to situations encoun-tered throughout the United States. The terminology presently used in Part 100 Appendix A is really applicable to western seismicity; its application to the Eastern U.S. is problematic. For a discussion of the difficulties encountered in applying Appendix A terminology to eastern seismicity, see Aggarwal and 10

Sykes, "Earthquakes, Faults, and Nuclear Power Plants in Southern New York and Northern New Jersey," Science, Vol. 200, pp. 425-429, April 28, 1978. B. Minimum SSE The proposed rule would continue the present practice of requir-ing the minimum SSE ground motion to be 0. lg. This is insuf-ficient. The ACRS has repeatedly recommended that the minimum_. SSE be raised to 0.2g. Okrent, p. 282 *(citing a*June 11, 1973 ACRS memo); Aggarwal and Sykes, supra, p. 428 (citing a January 17, 1977 ACRS letter). OCRE recommends that the NRC take the ACRS advice and raise the minimum SSE to 0.2g. Doing so would provide additional margin against the considerable uncertainties inherent in predicting seismic hazards. C. Use of Probabilistic vs. Deterministic Methods Upon review of the material supporting this rulemaking, OCRE believes that probabilistic methods can provide useful insights and have the ability to incorporate uncertainties. However, OCRE believes that the SSE should be established by the use of deter-ministic methods, and that the SSE should represent the maximum earthquake potential at the site, considering the regional and local geology and seismology and specific characteristics of local subsurface material (the Part 100 Appendix A definition of SSE). Nuclear power plants must be designed to withstand the maximum possible earthquake which could affect the site, based on a comprehensive study of the regional geology, regardless of its probability of occurrence. A large earthquake with a low annual probability (e.g., a quake with a 10,000 (+ or 1000) year return time has an average annual probability of l0E-4) could occur today and wreak havoc; if its last occurrence was 10K years ago, we have no way of knowing exactly when its recurrence is due. Its occurrence today does not make it more probable, as it won't recur for another l0K years. Low probability does not mean impossible. The SSE must be the maximum possible earthquake ground motion at the nuclear plant site, regardless of its proba-bility of occurrence. Proposed Part 100 Appendix B, Section V. (c), includes the re-quirement that "the annual probability of exceeding the Safe Shutdown Earthquake Ground Motion is considered acceptably low if it is less than the median annual probability computed from the current [EFFECTIVE DATE OF THE FINAL RULE] population of nuclear power plants." This appears to assume that the seismic risk to the current reactor population is acceptable. The fact is, for at least one operating plant, the seismic design is inadequate. 11

See SeismicitY ang_ Tectonic Structure in Northeastern Ohio: Implications ~ Earthquake Hazard to the Perry Nuclear Power Plant, by Dr. Yash P. Aggarwal, March 1987. In this report Dr. Aggarwal concludes that the seismic hazard in Northeast Ohio is greater than that assumed by the NRC, and that the Perry SSE does not provide the margin of safety required for nuclear power plants. Even if the probability of exceeding the SSE for the current reactor population (112) were acceptable, would that same level be acceptable £or a population of 1000 reactors? To maintain a constant level of cumulative risk to society from nuclear power, as the reactor population grows, the risk per reactor must de-crease. The proposed rule does not recognize that fact .

• III. Responses to Questions Posed in the Federal Register Notice OCRE is responding here to those questions (only for reactor siting criteria, nonseismic) which have not been addressed in the comments above.

1. Should the NRC grandfather existing reactor sites with an EAB leas than 0.4 miles for the possible placement of additional units?

Answer: Since OCRE believes that the EAB minimum radius should be expanded to 1 mile, it is not appropriate to grandfather existing sites for the placement of additional units. However, an appli-cant that wishes to add additional reactors to an existing site with an EAB less than the new standard can always seek an exemp-tion from the regulation.

2. Should the EAB be smaller than 0.4 miles for plants having reactor power levels significantly less than 3800 MWt and should the exclusion area be allowed to vary according to power level with a minimum value (for example, 0.25 miles)?

Answer: As explained above, OCRE believes that the EAB serves two functions: protection of the public from the reactor, and protection of the reactor from saboteurs. Thus, a minimum EAB of 1.0 mile is appropriate for all power levels for safeguards reasons.

6. What continuing regulatory significance should the safety requirements in 10 CFR 100 have after granting the initial oper-12

ating license or combined OL under Part 527 Answer: These safety requirements are part of the current li-censing basis. Since the main logical underpinning of the NRC's license renewal rule, 10 CFR 54, is the adequacy of the current licensing basis, which is assumed to be diligently maintained by the NRC, the NRC must in fact maintain this basis and must ensure that the safety margins of that basis are not eroded by changes in offsite population density and industrial hazards. IV. Misc. Mattera A. Typographical Errors OCRE would note the following typographical errors in the Federal Register notice:

p. 47804, second column, second full paragraph, 4th line, "Safety Boards Policy" should be "Safety Goals Policy"

p. 47804, second column, first paragraph under 2. Low Population Zone, 8th line, the word "belief" should be "behalf"

p. 47814, third column, footnote 7, let line, 11 24 rem" should be "25 rem" B. Submission of Comments in Electronic Format It would be more convenient to commenters if the NRC would estab-lish a BBS where comm.enters on all proposed rules (and other matters on which the NRC has solicited public comment) could upload the text files of their comments (including the use of file compression, such as ZIP format). This would avoid delays and the potential for damage to the magnetic media in transit through the mail. From an environmentalist's perspective, it would also be much less wasteful than using a 720K floppy disk to transmit a 30K file.

Respectfully submitted, Susan L. Hiatt, Director, OCRE 8275 Munson Road Mentor, OH 44060-2406 (216) 255-3158 13

MONTANA BUREAU OF MINES AND GEOLOGY *93 HAR 24 P4 :33 MONTANA COLLEGE OF MINERAL SCIENCE AND TECHNOLOGY BUTTE, MONTANA 59701 (406) 496-4180 Office of the Director March 15, 1993 Dr. Andrew J. Murphy, Chief U.S. Nuclear Regulatory Commission structural & Seismic Engineering Branch Division of Engineering Office of Nuclear Regulatory Research Washington, D.C. 20555

Dear Dr. Murphy:

I am enclosing comment by the Bureau staff seismologist on "Geologic and seismic siting criteria for nuclear power plants", Appendix A to 10CFR Part 100. sincerely yours, Geologist ETR/blm Enclosure MAY 111993 Acknowledged by card ....................., ~ THE BUREAU OF MINES AND GEOLOGY WAS ESTABLISHED BY LAW IN 1919AS A DEPARTMENT OF MONTANA COLLEGE OF MINERAL SCIENCE AND TECHNOLOGY, TO PROMOTE EFFICIENT DEVELOPMENT OF MONTANA'S MINERAL RESOURCES BY GATHERING AND PUBLISHING INFORMATION ON THE GEOLOGY, TOPOGRAPHY, AND MINERAL DEPOSITS OF THE STATE, INCLUDING METALS, NON-METALS, COAL, OIL, GAS, AND UNDERGROUND WATER SUPPLY.

1 S. NUCI E.' ,. - : :, " *~r.~ ( J'.,l\11SSIO~ DO"'l. -:- I '* '. .: **,-::r, i 10, J O I'* .ARY

                                         ,    I

-~... --_.......,....., . ,__ , . _. ___....... ~- "" l,-- I have reviewed the Nuclear Regulatory Commission' a <NRC > revised *Geologic and Seismic Siting Criteria for Nuclear Power Plants*, Appendix A to 10CFR Part 100. The stated purpose of th* revision is to *make it more consistent with current state-of-the-art and more flexible to changes in the earth sciences*. Appendix 0 of Draft Regulatory Guide 0G-1015 is entitled

                             *Geological,               Seismological and Geophysical Investigations to Characterize Seismic Sources*.                 Appendix 0 is a good review of methods useful for identifying both seismogenic and capable tectonic sources and determining their potential for generating earthquakes.               A total of 42 references are cited in appendix Das being pertinent to th* goal of identifying seismic sources and their seismogenic potential.               I note with some concern that recent works in the published literature are largely absent from the cited references.                 The accompanying histogram shows the year of publication for each cited reference.                 There seems to be little attention paid to work published since 1985.                 "uch good work has been published in the last six years and I feel that a document attempting to be *current with state-of-the-art* should include this recent work.

U) 8 I.al 0 z 7 til a: 6 I.al ta. I.al 5 a: Ca. 4 0 0: 3 Cal m 2 J:

J z 1 Y E A R Histogram showing year of cited references in Appendix 0 of Draft Regulatory Guide DG-1015.

DOCKET NUMBER Pl . f"h,,C_'_!'J PROPOSED RULE,..!.!!~ ~.-. C51 r-~47 iO 'J-) (7590-01]

                                                          *93 HA 23 A9 :57 NUCLEAR REGULATORY COMMISSION 10 CFR Parts 50, 52 and 100 RIN 3150-A093 Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Proposed Denial of Petition from Free Environment, Inc. et. al.

AGENCY: Nuclear Regulatory Commission. ACTION: Proposed rule: Extension of comment period.

SUMMARY

On October 20, 1992, (57 FR 47802) the NRC published for public comment a proposed rule to update the criteria used in decisions regarding power reactor siting, including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. The coment period for this proposed rule was to have expired on February 17, 1993.

On January 5, 1993 the public comment period was extended to March 24, 1993 (58 FR 271). The Commission has received a request to once again extend the public comment period based on the fact that the proposed rule presents difficult issues requiring thoughtful and careful analysis if the comments are to be of maximum value to the Commission. In particular, preparation of such comments involves careful consideration of the interplay between the proposed '3 f1-b l'fJ p~,at"-

                                                                        ~ SB'f fl 1'"~11

demographic and seismic criteria and the relationship of the proposed criteria to the Co11111ission's Safety Goals, severe accident requirements, and 10 CFR Part 52, as well as preparation of supporting analyses . The Co11111ission therefore finds that it is reasonable to extend the public co11111ent period to June 1, 1993, in order to allow all interested persons adequate time for such consideration. DATE: The co11111ent period has been extended and now expires June 1, 1993. Co11111ents received after this date will be considered if it is practical to do so, but the Co11111ission is able to assure consideration only for co11111ents received on or before this date. ADDRESSES: Mail written conaents to: Secretary, U.S. Nuclear Regulatory Co11111ission, Washington, DC 20555, Attention: Docketing and Service Branch. Deliver co11111ents to 11555 Rockville Pike, Rockville, Maryland, between 7:45 am and 4:15 pm, Federal workdays. Copies of the regulatory analysis, the environmental assessment and finding of no significant impact, and co11111ents received may be examined at the NRC Publ ic Document Room at 2120 L Street NW. (Lower Level), Washington, DC. FOR FURTHER INFORMATION CONTACT: Or. Andrew J . Murphy, Office of Nuclear Regulatory Research, U.S : Nuclear Regulatory Commission, Washington, DC 20555, telephone (301) 492-3860, concerning the seismic and earthquake engineering 2

aspects and Mr. Michael T. Jamgochian, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Convnission, Washington, DC 20555, telephone (301) 492-3918, concerning other siting aspects. Dated at Rockville, Maryland, this 1,,vd day of March, 1993. For the Nuclear Regulatory Commission. Samuel J. Chilk, Secretary of the Commission 3

DOCKET NUMBER PR ENBL PHOPOSED RULE..::...;;.;__....__,,,__ {_ 5'? Fll'i,'502-) DIREZIO NE STUD/ E Jl/C:,JRC:HE Roma, L U C Secretary *93 H R 23 p 4 :13 U.S. Nuclear Regulatory Commission Washington, DC 20555

                                                                                                    , ljl,1,

• I

{Rif.: VDN23393/U44/2860) Attention: Docketing and Service Branch

Subject:

Proposed rulemaking 10CFR Pans 50.52. and 100, "Reactor Siting Criteria" (Federal Register Vol. 57, No 20.1 - October 20, 1992)

Dear Sirs,

we understand that on the subject proposed rule you would he interested in rt:<:civing -:ommc1Hs also from foreign utilities. We know that other European utilities have already written to you on similar concerns and we agree in general with their l:011111\.cnts, I lowcvcr, since we recognize;:J from tht: early sixlies the important role of the US technolo1;.'Y and regulations on the nuclear energy development in our <:ountry, we appreciate this opportunity and arc willing to provide directly the following comm1.~nts. ENEL spa closed all its three operating nuclear plants after the Chernobyl accident 011 Governmental decision, hut it is committed at the same time to leave open the possihility of using nuclear energy for future electricity needs and, therefor~. it is working with l JSA as well as with European partners to develop advanced reactor concepts, rcspondjng to improved safety criteria. Our efforts arc also aimed at reaching consensus at international levels for design and licensing criteria of a new generation of plants more "environmental friendly" and with distinc:tivc improved safety foatures; we arc convinced that international consensus on homogeneous principles would be very important from technical, commercial and puhlic perception standpoints. The most important design target for us h,L".i been from tho h6{;inning to in~lude in ih~ d~t:ign prm;t:t,i; ~11 th~ ,:,;;,n*:~i*nbl,;- sr:vr:r~ ~~,:-irl,,.nh: ~n,l, consequently, to reduce environmental impact below any significant effect in all plant conditions <."Onsidered in the design. We have c.liscussod an<l we are continuing to discuss with our internatiom1I partners safety improvements for the "next generation" plants and, in our opinion. for example, significant results h,1ve been already achieve<l hy ALWRs designs in the field of severe acciuent protection and minimization of environmental consequences. A consequential action and a tangible proof of hetter safety for the public could be a rcdu<.*cu need of protective actions in case of accidents and, in particular, avoidance of public evacuation in the short and even in the long terms. 11,esc goals have heen a significant incentive to develop new or improved engineered features and, finally, we hope, a step forward in safety anc.l puhlic protection. IAY 111993 Acknowledged by card ...................:**"....~:;;

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Having explained in the extreme summary our situation, we would like to provide comments on a specific part of the subject rule, and specifically the detailed quantitative population densities in :1 large area around the plant. We would like to draw to your attention the following points. 0 The Emergency Planning should be correlated with the huilt-in safety level of the reactor designs: a complete decoupling between the siting criteria and plant design (e.g.: without. considering the safety features and the size of the plant) woul<l reduce or eliminate any incentives to the reactor designers/utilities for searching additional safety improvements. 0 The future application of the proposed rule to new reactor plant designs would create an evident contraddiction in puhlic perception. New reactors, claiming improved s:ifety performances, would he subjected to an overprescriptive siting process: this would lea<l to helieve that new reactor plants respondt actually, to lower safety standards.

• 0 o

For the "next 1Jt'.n1~r:ition" /\1 .WRs, even long term protective actions, if nee<le<l at all, woul<l he limited to a small arc.a much closer than JO miles radius mcntione<l in the proposed rule. If we apply the proposed rule, it would he almost imposi;ihlc (or at. least very difficult) in Italy. as well as in many other European and Ac;ian countries, to find any sites complying with the new rule; in any case the choice of the site would be certainly driven by demographic consideration, giving a reduced priority to other

         -.afoty con~iderations such as seismic level an<l meteorological conditions.

o We agree that t>)(trapolation of popul~tion <lcnsitics to 40 years after initial plant operation i~ vt:-ry unrdiable and, in addition, it is not clear how this could fit with the plant design life of sixty yeari- proposed for next generation reactors. n Tf quantitative population dcnsiLies could not be simply deleted from the proposed rule, we suggest to put them only in guidance documents, such as Regulatory Guides, conceive<l as preferre<l ways of complying with the regulations, and, therefore~ more nexible. We hope that these cnn11ne11ts could he taken into consideration, so as to avoide the envisaged negative effects to future ._fovclopment of nuclear energy in Italy as well as in other countries. Sincerely yours l ngg. Velona ~ Fornaciari IT/Id

ENEL S.p.A. - DIREZIONE STUDI E RICERCHE Research & Development Division Via 0 .8 . Martini 3

• 00198 Rl)ma Pholl'l( + 39 6) 85091 Telex: 610518 ENEL OG VICE DIREZIONE NUCLEARE Nuclear Departmenc Viale Regina Margherita 137 00198 Roma Telex: 6l05t8 ENEL DG Phone : ( + 39 6) 85391 Telefax : ( + 39 6) 85398601

( ~39 6) 85091 +39 6) 85098601 Address : To: Se<.:retary LJ.S . Nuclear Regulatory Commission Fa.x N.: 00 I 30 l 504 2260 Sender : From : Ing. P. Fornaciari ENEL DSR/VDN Subject : Proposed rulemaking I0CFR Parts 50.52, and 100, "Reactor Siting Criteria"

• (1--cdcral Register Vol. 57, No 203 - October 20, 1992).

Dear Sirs . please find hereafter a (.'.opy of lhc letter with the TINEI . comments Lo Lhc Prnpust!d rulemakrng LOCf-R Parts 50 ,51 . ,-tnd 100, "RcacLor Siting Criteria" (Federal Register Vol. 57, No 203 - ()ctoher 20, 1992) .

• Du~ to the importance of the issue for the future tlevelupment. of nuclear energy in Italy as well as in olher countries, we are thankful fnr the opportunity to provide you our view on the suhject.

Sincerely yours 1 ' Fornaciari Paplo I I I .I ).f /Id Transmission date : Hour : Sheets number: Signature: 1096£98 90 6£,S2.

 ~   TRANSACTION REPORT>                        03-23-1993(TLJE) 11:48 C   RECEIVE        J NO. DATE  Tlt-E  DESTINATION STATION PG. DURATION   MODE     RESULT 10385  3-23 11:45  39 06 8539601         3 0"02 ' 49" NORMAL   OK 3 0*02*49*

I

DOCKET NUMBER F;*,._ ;=-oSED RULE PR ~ . ~ 5 v *-

• tJ--/00 G C Sla11is Associates C£ 1 FIL ~ 7 s-oa)

CONSULTING ENGINEERING 3520 Eris Court

• Walnut Creek
• California 94598-4669 * (510) 943-7101

                                                                 *93 MAP- 23 p t: :2' March 22, 1993                                                      I   l.,r G93-03 Secretary U.S. Nuclear Regulatory Commission Washington, DC 20555 Comments on Proposed Rules FR, V57, No. 203, Tuesday, October 20, 1992 Appendix S to Part 50 Appendix B to Part 100 My major comments are on the single earthquake design approach. My perspective is that of a person who has been designing piping systems, including seismic effects, since 1971.

At present, I chair the Working Group on Piping Design (SC 111). This Group is responsible for nuclear piping design rules of the ASME Boiler and Pressure Vessel Code. Design for a single limiting event and inspection for lessor earthquakes is not a sound regulatory approach for pressure-retaining components designed to Section III. There are several errors of reasoning in the justification for "SSE only" design. Adopting this approach will severely compromised safety. The two earthquake criterion is a prudent way to insure the safety of a nuclear power plant. Newmark proposed this approach in the 60s. The OBE was originally called the design earthquake, and the SSE, the maximum credible earthquake. Since the OBE was expected to occur in the life of the plant, you "designed " for it using normal allowable stress limits. Since the SSE was not expected to occur (low probability), you didn't "design" for it using normal allowable stress limits. Newmark did not specify the acceptance criteria. Severe deformations were expected, and decisions on margins of safety would have to be made based on the nature and importance of the structure. Pressure-retaining components in the nuclear side of a plant have to meet ASME Section III requirements. Section III provides stress criteria for Design, Level A (normal), Level B (upset), Level C (emergency) and Level D (faulted). OBE is a Level B condition. The Level B stress criteria ensure that the pressure retaining component can withstand the loading without damage requiring repair. Cyclic considerations (fatigue) are included. SSE is a Level D condition -- an extremely low probability event. The stress limits for Level D are much higher than Level B, and cyclic (?ffects are not considered. The component will survive the loading, but there may \~ gross structural damage requiring f/lY 111993 Acknowledged bv Mttf ..............-...- ....,_.::;.

lJ.S NUt:- 1.:::*;-, *e~ jl \11:--1'.' COMMISSIO~ DO '.:. -. - :,~ r '.S:: S'.::CTION 0 , .,-;*. > :, *~: e.~r*1::T,\RY o, ;11- ~c:,., ..::..:,;:Jr J

G C Sla11is Associates March 22, 1993 G93-03 Page Two replacement of the component. Section III criteria are consistent with the Newmark approach. Design for the expected earthquake, where design means no damage, and accept large deformations but survive the extremely low probability earthquake. The key to understanding the Section III requirements is that cyclic effects are controlled for Level B, but not for Level D. In the discussion of the proposed rules it says, ". . . the NRC staff states that it agrees that the OBE should not control the design of safety systems. " [V.B.5] I cannot understand how the NRC staff can make this statement. If the site characteristics are such that significant lower level earthquakes are expected during the life of a plant, then cyclic effects may be significant. In this situation, cyclic effects could control the design. Therefore, OBE controlling the design is appropriate, and necessary, in some situations. We do have a problem in the industry with the present requirements. Requiring "design" for five OBE events at 1/2 SSE is unrealistic for most (all?) sites and requires an excessive and unnecessary number of seismic supports. The solution is to define appropriately the OBE magnitude and the number of events expected during the life of the plant. And to require "design" for that loading. OBE may or may not control the design. But you cannot assume, before you have the seismicity defined and before you have a piping system design, that OBE will not govern the design. Of course, you can ensure that OBE will not control design by arbitrarily defining very limiting SSE stress criteria. But, this is not a reasonable approach since it would require too many pipe supports. Implicit in the reasoning behind the proposed rules is that if a piping system (or other pressure-retaining component) meets the Section III requirements for an SSE as a Level D condition, then that piping system (or other pressure-retaining component) will automatically satisfy Level B stress criteria at 1/3 SSE. Obviously, if a piping system ( or other pressure-retaining component) can survive an SSE, then that component can survive an OBE at 1/3 SSE. That is not the technical issue. The technical issue is whether significant cyclic fatigue "usage" will occur. Fatigue usage from the OBE reduces the cyclic life for the other Level A and B conditions. Without explicit consideration of the earthquake cyclic stresses, these stresses would have to be below the endurance limit of the material to have no influence on the cyclic life. This is highly improbable. Therefore, to say that, 'The proposed regulation would allow the value of the OBE to be set at: (i) One-third or less of the SSE, where OBE requirements are satisfied without an explicit response or design analysis being perfonned, ... " [V.B.5], is not technically justified. I cannot understand how this decision was reached. I can imagine that if you looked at some piping systems designed to the present requirements, you may be able to show that OBE seismic effects are insignificant. But this is not relevant. Systems designed to the new rulemaking, without design for OBE at 1/3 SSE, would be very different. There is no way, as far as I can see, to make assumptions about the earthquake stresses at 1/3 SSE if you do not design for the OBE loading.

G C Sla11is Associates March 22, 1993 G93-03 Page Three The problem with not designing for OBE can be simply stated. The piping ( or other pressure-retaining component) may be designed to the limit for other Level A and B loads (for example, thermal expansion cycling). In this situation, OBE stresses above the endurance limit reduce the operational life of the component. It is highly improbable that OBE stresses will be below the endurance limit. The only way to accept the OBE stress cycles is to accept lower margins of safety. This is compromising the design of the plant, and is unnecessary. Design for OBE, if the OBE magnitude is reasonably defined, will not result in an excessive number of seismic supports. The error in the logic of not requiring design for QBE is evident in the last statement in V.B.5., With regard to piping analysis, positions on fatigue ratcheting and seismic anchor motion are being developed and will be issued for public comment in a draft regulatory guide separate from this rule making. " If you understand piping design, you realize that this statement means that it is not valid to assume that the OBE requirements (at 1/3 SSE) are satisfied without an explicit response or design analysis being performed. What this statement implies to me is that NRC is going to specify stress criteria in a regulatory guide. This is inappropriate! As a member of the Section III code committee, I object strongly to NRC defining stress criteria. Stress criteria should be the responsibility of the ASME Boiler and Pressure Vessel Code committee. What I expect to happen, if the proposed rules are implemented, is that NRC will require fatigue analysis for the SSE and the OBE events. This essentially means considering SSE as a Level B condition. This is too conservative, unreasonable, and will require even more pipe supports than the present regulations -- a step in the wrong direction. The first error is the assumption that a pressure-retaining component automatically satisfies the OBE requirements (at 1/3 SSE) if the SSE requirements are satisfied. The second error is the assumption that a utility will be able, by inspection and test, " ... to demonstrate to the Commission that no functional damage has occu"ed to those features necessary for continued operation without undue risk to the health and safety of the public." [V.B.6] If an earthquake at slightly above 1/3 SSE occurs, the plant has to shut down. The piping systems (and other pressure-retaining components) have not had explicit response or design analysis performed. It is not feasible, by inspection or test, to decide whether the earthquake impacts the cyclic life of the component. Obviously, you will be able to tell if the pressure boundary is leaking, but that is not the issue. The issue is whether the earthquake has "used up" an unacceptable amount of the cyclic fatigue life. The only practical way to assess fatigue usage is by analysis. A piping system is a collection of many fittings and joints. The maximum stressed locations in the system and within the individual fittings and joints are not readily determined. I cannot see how you can determine the amount of cyclic life used by the earthquake with inspection or test. The third error is that some pressure-retaining components required for plant operation after an earthquake will not have explicit response or design analysis for the OBE or the SSE. The three types of structures, systems, and components that must be designed to

G C Sla11is Associates March 22, 1993 G93-03 Page Four remain functional for the SSE are a subset of the structures, systems, and components necessary for continued operation without undue risk to the health and safety of the public" (Appendix S, III]. Therefore, the structures, systems, and components that are not required for shutdown, but are required for safe continued operation, will have no seismic qualification performed. This certainly is not prudent. In summary, design for "SSE only" is not a prudent approach, and the safety of a nuclear plant will be severely compromised by this approach. It is not technically valid to assume that a Section III pressure retaining component that meets SSE requirements automatically satisfies the QBE requirements. It is not technically valid to assume no impact on the cyclic life of a Section III component if an QBE at 1/3 SSE occurs. It is not technically valid to rely on inspection or test after an QBE event to determine whether the QBE event has reduced the cyclic life of a component. It is not prudent to require no seismic qualification for structures, systems, and components that are not required for safe shutdown but are required for continued operation. The intent of the rule making, to uncouple the QBE and the SSE, is a necessary change in the seismic requirements. The problem is that the proposed rules are not valid. There is a simple solution. My recommendations follow. Appendix S should define the magnitude of both the SSE and QBE ground motion and require design for both earthquakes. I see no reason to arbitrarily set the QBE at 1/3 SSE instead of the present 1/2 SSE. My preliminary suggestion is to set the SSE at lE-5 to lE-6 and the QBE at lE-1 to lE-2. There may be certain structures or systems for which a separate analysis for QBE is not required to verify the seismic capability. A regulatory guide can be prepared to specify under what conditions only one seismic analysis is needed. I appreciate the opportunity to comment on the rulemaking, and I hope my comments are clearly understood. We need to be able to design more cost effective piping systems. To uncouple the QBE from the SSE is an appropriate way to allow practical and safe design of piping systems. To not require design for QBE is not prudent. I do not see how we can explain to the public that we do not need to design a nuclear plant for the earthquake loading that we expect to occur in the life of the plant.

~h Ge~       . Slagis

DOCKET NUMBER

                                         , ~ " " ). ED RULE PR S Z:,,S 1- J-/ O O        *

(§7 FP.'/7 f01-) * ~ @ Your ref Scottish Our ref Lc10FM Nuclear AIRMAIL Scottish Nuclear Limited 3 Redwood Crescent Mr S.J. Chilk Peel Park East Kilbride G74 5PR Secretary to the US Nuclear Regulatory Commission Fax (03552) 62626 WASHINGTON DC 20555 Telephone (03552) 62000 U.S.A . 22 March 1993 MAR 2 3 993 D0cKETINQ& IEfMCEIIWICJN SECY-HRC

Dear Mr Chilk,

PROPOSED REVISION OF THE RULES RELATING TO SITING CRITERIA You will appreciate that any increase in the stringency of siting criteria for nuclear power stations in the US is likely to be used by intervenor groups in the UK to secure a similar increase in this country. Accordingly, we have studied carefully the proposals announced in the Federal Register on 27 October 1992, together with the accompanying discussion, and have responded fully to the Commission's request for responses to the set of 8 questions. I trust that you will find these responses useful. Yours sincere! y, R.J. KILLICK Director of Safety & Quality Acknowledged by card ....."...!.!..!JJ........"": LCL04033.001 Registered in Scotland Number 117121 Registered Office: 3 Redwood Crescent, Peel Park, East Kilbride G74 5PR VAT Number 552 2693 40

U.S. l\itir.* * ~tGULATOJ:tY COMMISSIO,-, DO,_*. ; ,~G & SEPMCE SECTION Or;.tCE OF THE SECRETARV Of THE COMMISSION DoOJment Slltistics Postmark Dall ?; I ;z_' 11.J /=b.J-ij ~ J/2.J /7 3 Copies Reeeived 1 Add'I Copies Aeproduc.'td L ] _____ _ Special ctstributlon rL:rYif; Pa'1/2 . /1?<- c:J/lh y..) J a.~~J c:. IA1 ~ I,'\.

RESPONSES TO THE COMMISSION'S QUESTIONS SECTION A REACTOR SITING CRITERIA {NON-SEISMIC)

1. Should the Commission grandfather existing reactor sites having an exclusion area distance less than 0.4 miles (640 meters) for the possible placement of additional units, if those sites are found suitable from safety consideration?

Response

Yes. - 2. Should the exclusion area distance be smaller than 0.4 miles (640 meters) for plants having reactor power levels significantly less than 3800 Megawatts (thermal) and should the exclusion area distance be allowed to vary according to power level with a minimum value (for example, 0.25 miles or 400 meters)?

Response

Yes, the exclusion area distance should be smaller for lower power reactors; the possibility should be kept in mind that, if small modular reactors with a high degree of intrinsic safety are introduced, it might be possible to eliminate the need for an exclusion area other than that provided by the security fence for the Station.

3. The Commission proposes to codify the population density guidelines in Regulatory Guide 4. 7 which state that the population density should not exceed 500 people per square mile out to a distance of 30 miles at the time of site approval and 1000 people per square mile 40 years thereafter. Comments are specifically requested on Questions 3A, 3B and 3C given below.

Responses: 3A. Should numerical values of population density appear in the regulation or should the regulation provide merely general guidance, with numerical values provided in a regulatory guide? Scottish Nuclear Limited would much prefer that numerical values should NOT appear in the regulation. We consider that this is consistent with the discussion of the Commission's rationale for this change in the regulations which is provided on pages 47804, 5 and 6 of the Federal Register. 3B. Assuming numerical values are to be codified, are the values of 500 persons per square mile at the time of site approval and 1000 persons per square mile 40 years thereafter

appropriate? If not, what other numerical values should be codified and what is the basis for these values? Scottish Nuclear Limited does not consider that the proposed values are appropriate for future reactors. The probability of a severe accident should be at least a factor of ten less than that for the reactors designed prior to 1975. As the proposed population densities out to 30 miles appear to have been derived from an analysis of the risks from these earlier designs, it is unlikely that they will be appropriate for future designs. As discussed in the response to Question C, it is the distance to which population densities might be limited that is important, not the density out to the proposed distance of 30 miles. 3C. Should population density criteria be specified out to a distance other than 30 miles (50 km), for example, 20 miles (32 km)? If a different distance is recommended, what is its basis? For the same reasons that the Commission has suggested that the exclusion distance could be varied with reactor power, the population density criteria should also be varied in the same way. Scottish Nuclear Limited does not consider it appropriate to make any specific recommendations for US conditions but, as indicated above, it considers that the Commission should review the analysis carried out prior to 1990, in the light of the higher standard of reactor safety which is expected to be provided in future Stations.

4. Should the Commission approve sites that exceed the proposed population values of 10 CFR 100.21, and if so, under what conditions?

Response

As discussed in the response to Question 2, Scottish Nuclear Limited does not consider the proposed population values are appropriate. Assuming that the Commission will put forward a more appropriate set, there should still be some flexibility in their application, as indicated in the Commission's discussion of its rationale for the proposed change. Approval of sites that do not meet the population criteria should be considered on a case-by-case basis.

5. Should holders of early site permits, construction permits, and operating licence permits be required to periodically report changes in potential off-site hazards (for example, every 5 years within 5 miles)? If so, what regulatory purpose would such reporting requirements serve?

Response

2

This problem would not arise in the UK as the regulatory body (The Nuclear Installations Inspectorate, NII) has to be consulted if proposed developments in the vicinity of a licensed nuclear site are likely to present an additional hazard to the Station. The NII can recommend to the local Planning Authority that the proposal should be amended or rejected; it is most unlikely that their advice would be rejected. It is not entirely clear from the discussion of Recommendation 5 that the NRC has comparable powers in this respect. However, if planning controls in the US do not give the Commission these powers, there could be numerous opportunities for intervenor action immediately prior to fuel loading of a reactor licensed under the combined operating licence procedure of 10 CFR 52.

6. What continuing regulatory significance should the safety requirements in 10 CFR part 100 have after granting the initial operating licence or combined operating licence under 10 CFR part 52?

Response

Scottish Nuclear Limited considers that the regulatory significance of these safety requirements would continue unchanged after the operating licence had been granted.

7. Are there certain site meteorological conditions that should preclude the siting of a nuclear power plant? If so, what are the conditions that cannot be adequately compensated for by design features?

Res_ponse The accident situations in which meteorological conditions would be important are outside the design basis. If the local conditions are much more adverse than average (e.g. a Station sited in a valley deep enough to distort severely the wind pattern) the only design action that could be taken would be to improve the safety systems further, in order to reduce the probability of the accidents occurring.

8. In the description of the disposition of the recommendations of the Siting Policy Task Force report (NUREG-0625), it was noted that the Commission was not adopting every element of each recommendation. Are there compelling reasons to reconsider any recommendation not adopted and, if so, what are the bases for reconsideration?

Response

As discussed above in relation to Question 5, it may be appropriate for the Commission to review the powers which it has currently to control development that could present a hazard to a licensed nuclear installation and, if necessary, seek stronger powers. 3

SECTION B As none of the proposals in this part of the proposed rule conflicts with current UK practice, no comments are offered. LCN04033.001 4

123 Main Street DOCKET NUMBER PR ()IJ White Plains, New York 10601 PROPOSED RULE .- f 914 681.6846 [G--t FR '118-0:J.-)

     , . NewYorkPower                                                                                        Ralph E. Beedle
     -1 Authority                                                    *93 W'1 22 P'1 :13                      Executive Vice President Nuclear Generation I    ,hl                       March 18, 1993 IPN-93-013 JPN-93-018 Secretary of the Commission U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Docketing and Service Branch

Subject:

Indian Point 3 Nuclear Power Plant Docket No. 50-286 James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Comments on Proposed Rule 10 CFR Part 50, 52, and 1oo: Reactor Site Criteria

References:

1. NRC Notice of Proposed Rulemaking, Volume 57, Federal Register 47802, dated October 20, 1992.

2. NRC Notice of Proposed Rulemaking, Volume 57, Federal Register 55601 ,

dated November 25, 1992.

3. NYPA letter, R. E. Beedle to Chief, Regulatory Publications Branch, U.S.

NRC, dated May 4, 1992 (IPN-92-021/JPN-92-019}, regarding Comments on Elimination of Requirements Marginal to Safety.

Dear Sir:

The New York Power Authority has reviewed the proposed changes to Title 10 of the Code of Federal Regulations, Parts 50, 52, and 100 concerning "Reactor Site Criteria" (Reference 1). The Authority also reviewed the related draft Regulatory Guides and Standard Review Plan Section 2.5.2 (Reference 2). The Nuclear Management and Resource Council, Inc. (NUMARC) has reviewed the proposed rule and regulatory guides and developed general and specific comments. This letter endorses the NU MARC comments and provides additional specific comments for your consideration. The proposed rule is overly conservative and would unduly restrict future siting possibilities at one Power Authority site with a currently operating nuclear power plant. This site meets all current regulatory requirements for licensing. This was confirmed not only at the time the construction permit and operating license were granted, but also after an exhaustive special adjudicatory proceeding concerning a determination of the extent to which the population around the Indian Point site affects the risk posed by an accident at that plant. A discussion of the MAY 1 1 1993- -. Acknowledged by card .................................

.s. r, * -:. . . "S:J [* ' . r- ** I

2 proceeding with conclusions are contained in Commission decision dated May 7, 1985 (CLl-85-6, 21 NRC 1043). Despite the fact that the NRC has determined this site to be safe, it could not meet the exclusion area or population density requirements in the proposed rule. This could call into question the current operating license and would preclude the use of this site for a new plant when the current operating plant is decommissioned. The decommissioning of an existing plant, its removal, and future use of that site for a new plant, should not be excluded because of an inflexible rule based on more limiting population criteria. As a minimum, the rule should allow the use of previously accepted sites in accordance with the original regulatory criteria. The present regulations, specified in 10 CFR 100, provide sufficient guidance and protection in the event of a reactor accident. This has been demonstrated, even for greater than design basis accidents, in a report titled "Graded Response: The Preferred Evacuation Strategy For Nuclear Power Plants" (NUMARC/NESP-005, February 1989). This report and the Authority's position were provided in the Authority's May 4, 1992 letter (Reference 3). The Authority also recommends the greater use of Probabilistic Risk Assessment (PAA) as a basis for regulations, specifically regarding seismic regulations, rather than the present deterministic basis. We request that the present requirements in 10 CFR 100 be retained and that the proposed regulations be abandoned. This action will provide the necessary protection of the public while maintaining the necessary flexibility in site selection for nuclear power plants. If you have any questions, please contact Mr. P. Kokolakis or Mr. J. A. Gray, Jr..

                                                                   *~

Ralph E. Beedle cc: U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector's Office Indian Point Unit 3 P.O. Box 337 Buchanan, NY 10511 Office of the Resident Inspector U.S. Nuclear Regulatory Commission P.O. Box 136 Lycoming, NY 13093

3 Mr. Nicola F. Conicella, Project Mgr. Project Directorate 1-1 Division of Reactor Projects - 1/11 U.S. Nuclear Regulatory Commission Mail Stop 14B2 Washington, DC 20555 Mr. Brian C. McCabe, Project Mgr. Project Directorate 1-1 Division of Reactor Projects - 1/11 U.S. Nuclear Regulatory Commission Mail Stop 14B2 Washington, DC 20555 Mr. Thomas E. Tipton NUMARC 1776 Eye Street, NW, Suite 300 Washington, DC 20006-24956

Vereinigung Deutscher E/ektrizitiits werke -VDEW-e.V. Postfach 701151 Secretary US Nucle a r Regu latory Stresemannal/ee 23 0-6000 Frankfurt/M. 70 Tele/on (0 69) 63 04-2 50 c!J/

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Washington DC 20555 Telefax (0 69) 63 04-3 68 Telex 411284 Biiro Bonn Attention: Docketing and Service Br anch Friedrich-WJ/helm-Stra8e 1 0-5300 Bonn I Tele/on (02 28) 2310 31 Telefax (02 28) 23 6160 Telex 886582 Datum March 1 8, 1993 Biiro Brussel - zetchen Az .: Kie/rk 250.1.0 148, Avenue de Tervuren, Bte. 17 B-1150 Brussel Tele/on (0 03 22) 71196 42 Telefax (00322) 7630817 Proposed revision to US siting regulation Federal Register Vol. 57, number 209

Dear Sirs,

the US NRC has issued for comments a proposed modifica-tion of 10 CFR 100, t he deadline for submitting com-inents being March 24, 1 993 . Recent cont acts wid1 the US NRC have shown t hat it would recei v e with interest com-ments from foreign organizations . Vereinigung Deutscher Elektrizit a tswerke (VDEW) acting on behalf of the German Nuclear Power Operators welcomes the opportunity thus afforded and would like to confirm the arguments already exchanged between Mr. Murley , NRC, and Mr . Kienle , Head of Nuclear Power Di-vision , on the occassion of the NRC/USC meeting held on January 22 , 1993 , in Palo Alto. You are kindly asked to take into account our arguments in a final version . Consistency between Plant Safety and Choice of Sites In 1981/82 the European and Japanese utilities had a l - ready stated their deep concern about the intention to lay down a rule for tolerable site conditions under discussion at that time . Main reason for the rejection of this former plan was our argumentation that site criteria cannot be defined independently of the design and safety goals of the plant to be built . Consequent-ly, NRC has turned to the definition of safety goals aiming at a significant improvement of plant safety compared to the presently existing plants. Oresdner Bank AG Frankfurt/ M. Bil 50080000 MAY 111993 Kto-Nr. 97191100 l.lUIUWIOOQEKJ oy Ca1u .....~..... -~.::ru. ~- ' Postgiro Frankfurt/ M.

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Thus the announcement of new and sharper safety re-quirements becomes even more incomprehensible to us, as the safety level of presently planned plants which shall be built after the second half of the nineties has significantly been increased especially due to the EPRI requirements. If these new plants, which shall receive a design certification until the end of 1995, have safety fea-tures limiting the impacts of an accident to the site boundary, it is incomprehensible why people are not allowed to settle within an exclusion zone beyond the plant fence and why the number of inhabitants is dra-stically limited within a radius up to 50 km. This li-mitation leads to the consequence that the construction of nuc le ar power p lants would only be permitted in a large distance to population centres. Thus the option to use these plants also for district heating purposes has to be directly excluded, although this option would be desirable due to reasons of energy policy and climate protection. Public Concern In our view it cannot be explained to the public why such inconsistent requirements are requested by the authorities. Moreover, there is the fear that the public might become suspicious that the allegedly safe new reactor generation really shows significant defi-ciencies with respect to the control of severe acci-dents. Otherwise such a preventive concept against ca - tastrophes would not be necessary. As roughly 25 % of the present US sites for nuclear power plants do not fulfill the new criteria, the existing plants on these sites would require an excep-tional permission or the schizophrenic case has to be explained to an astonished public that - on the same site - more modern and safer reactors are not allowed to be built, whereas the older reactors with lower safety levels exploit the residual lifetime of their operating license. International Solidarity Apart from these general reasons, we want to draw your attention to our special national concerns that practi-cally none of our present sites is able to fulfill the criteria you demand. Due to data available to us, the

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I I I I same holds for many sites in Great Britain, France and Japan. With regard to the International Community of States operating nuclear power plants we appeal to you not to publish any regulations (if not stringently required due to a specific national site situation) which might unnecessarily aggravate the controversially discussed use of nuclear power in many countries. The intentions pursued by NRC according to Mr. Murley (avoidance of plant sites within centres of dense population) could be fulfilled if population densities, which are compatible with site conditions in other countries, are accepted. Less disadvantageous for non - US states would ~lso be a procedure not to fix the intended data in a formal regulation but in a legally less binding administrative rule in the sense of an intended goal a nd not as a rigid boundary to be fulfilled in any case. Preservation of the German Option for Nuclear Power You are aware of the specific German situation where presently a hard strugg l e takes place for a political consensus concerning the future use of nuclear power and where new conditions may be fixed concerning licensing, construction and operation of new nuclear power plants. In this situation the German utilities are especially sensitive with respect to regulations, which they are unable to fulfill due to natural con-ditions. It cannot be excluded that a future use of nuclear power in our country will only be possible if we are willing to accept the worldwide highest safety requirements both concerning the plant design itself and the sites. Your intended new regulation might therefore significantly influence the long-term use of nuclear power in our country. Yours sincerely Vereinigung Deutscher Elektrizitatswerke - VDEW Prof. Dr. Joachim Grawe

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    ~Ul)l:C Nuclear Power Engineering Corporation I, ""    * -

, ,C Secretary *93 HAR 22 A10 :59 U.S.Nuclear Regulatory COU1Dissio11 Washington D.C.* 205fi5. U.S.A.

Attn: Docketing Service Branch Re: Proposed Rule on Reactor Site: Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants March 15th. 1993 Dear Sir. Your kind arrangement to send Dr.Congel a~d Mr.Soffer to OECD headquarter in Paris on January 14th to introduce the proposed rule on reactor site in detail is highly appreciated. I attended the meeting as a member of Japanese delegation and had learned that the proposed rule was based on defenth-in-depth concept and the same rule should he applied to nuclear sites independent of the plant design whi le in other countries there would be trade-off between the plant design and the siting to achieve safety. Responding to the staff's introduction at the meeting. l expressed my concern that the size of the exclusion area and the distance to the population center - appropriate for the U.S. might not be for other countries of OECD. H would be clear that local circumstances of various nature (political.societal. geograph-ic.etc.)played a dominant role in the mix of criteria adopted by each country. If the size of the exclusion area and the distance to the population center in the proposed rule are misunderstood as the minimum values to achieve safety for any country, it would give negative effect to the development and the restora-tion of nuclear power. Accordingly. I would like to ask your consideration of ma.king it totally clear in the new rule that ihose values may vary from country to country de-pending on their societal. geographical. and .Po litical con<lHions. Your consideration to the above is highly appreciated. ~y 111993

                                                            , n, edged by card ................................ ._
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Usisa_yoshi Shiba JAPAN IN:sTITUTE OF NUCLEAR SAFETY -"'~ ~~i!- Masa'Yoslii Shiba JITA KANKOU RA MON D~. 7F 3*17~1. TORANOMON Director General MINATO-KU, TOKYO. 105. JAPAN Insti tute of Nuclear Safety. NUPEC

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ELECTRICITE DE FR ANCE 32 , RUE DE MONCEAU - 75008 PARIS LE DIRECTEU R GE N ERAL J~- ,~4# TEL . (1 ) 42 67 94 10 ADJOI NT

                                   *93 MAR l,q P2 :5 1
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• t.,11 ,. , '\ I' '-! Secretary ui;l,kt , *, * 'v 1r,f US Nuclear Regulatory Commission

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Washington DC 20555 Attention : Docketing and Service Branch 10 "ARS \993 Proposed revision to US siting regulation Federal Register Vol. 57, number 209

Dear Sirs,

The US NRC has issued for comments a proposed modification of 10 CFR 100, the deadline for submitting comments being March 24, 1993. Recent contacts with the US NRC have shown that it would receive with interest comments from foreign organizations. Electricite de France welcomes the opportunity thus afforded. The US regulations always have had a deep influence on French regulations and practices. This has particularly been the case for regulations which have a direct impact on the public, such as the one on siting. Our comments will bear mainly on the proposed demographic criteria, but we shall also comment the proposed seismic criteria.

1. Demographic criteria a) The proposed demographic criteria (exclusion zone and population density in a 30 miles radius) are independent from the design of the nuclear power plant to be built on a given site. Although we think we understand the motives which underly such a decoupling between design criteria and site criteria, we believe this decoupling will send a wrong signal to the plant owners and designers, and will not encourage them to improve further the built-in safety margins in their future plants.

b) We believe that average population density is not a sufficient, nor a universal, demographic criterion. Areas with locally concentrated population, evacuation routes, prevailing wind conditions, and many other factors may have to be taken into consideration for emergency planning. This is why we feel that a regulatory guide is better suited than a rigid regulatory rule, inasmuch as it may be more flexible . MAY 111993 Acknowledged by card ....- ...................~::::; SIEGE SOCIAL : 2, RUE LOUIS M URAT

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• c) We understand that several existing US sites would not meet the proposed rule. This would also be the case for one site out of three in France, and probably the situation would be even worse in more densely populated countries both in Europe and in Asia. Our opinion is that existing sites, both in the US and in France, should be able to receive new plants with enhanced built-in safety characteristics and margins. If this were not the case, public acceptance of existing plants on these sites would become problematic.

To conclude on demographic criteria, we would suggest that the proposed rule be replaced by a regulatory guide (or a revision of the existing regulatory guides), that consistency be assured between past and future practice for site approval, and that some provision be introduced in the guide to allow flexibility -within certain limits-for higher population densities or simplified emergency planning if the built-in safety characteristics and margins of the new plant lead to lower offsite consequences in case of accident. We would also suggest that the optimum balance between demographic criteria and emergency planning should be identified as dependent on local conditions, since such an argument would support the flexibility we feel is necessary.

2. Seismic criteria a) Many probabilistic methods have been proposed for ground motion assessment, either in seismic building codes for ordinary buildings or in safety studies for critical facilities. Most of these methods follow the approach defined by Cornell in his pioneering work (1968), which is based on the construction of models for delimitation of earthquake sources, activity rates and magnitude distributions of these sources, and attenuation laws of ground motions A number of computer programs are now available for practical applications of these methods (Mc Guire, Mortgat, NOAA, Woodward-Clyde, Principia Mechanica, ... ).

The main problems encountered when applying probabilistic methods come from: 1/ the quality of data, 2/ the treatment of uncertainty, 3/ the use of expert opinion. These three points are also the key factors for deterministic analyses, but the way they are dealt with depends on the type of method : 1/ The basic seismic data are of course the same for both methods, but deterministic methods need fewer data than probabilistic ones ; for each earthquake source, deterministic methods essentially require the maximum expected magnitude and the closest (with respect to the site) location of the hypocenter to be estimated ; probabilistic methods need in addition the precise delimitation of the source and information on activity rates and magnitudes distribution. 2/ In any risk analysis, end results should be documented not as single numbers, but as probability distributions reflecting the uncertainty of these numbers. However the practice is somewhat different from the theory and the treatment of uncertainties is rarely tackled in depth in risk assessment studies. The virtue of deterministic methods is that the question of uncertainty is bluntly solved through the use of precise rules producing a single result, which is supposed to be safe enough. This may

                                                                                   .../ ...

obviously be criticised on scientific grounds, but it has the merit of simplicity. 3/ Many points in the analysis, either deterministic or probabilistic, rely heavily on expert opinion. The difficulties arising from the lack of scientific consensus on some of these points have motivated the search for formal means of eliciting expert judgement and aggregating individual opinions. These difficulties are perhaps greater for probabilistic methods. the results of which tend to be dominated by the tails rather than the central tendencies of the distribution of expert opinions. These comments, and especially point 2/ above, indicate that, in our opinion, probabilistic methods have not yet reached the state of a well established procedure for ground motion assessment in the licensing process of a nuclear power plant. This does not reflect a negative appraisal of the potentialities of these methods, but merely stems from the fact that precise guidance for their application, and particularly for the treatment of uncertainty, is not available at the present time. b) We have some doubts as to the wisdom of requiring two different methods for the safety assessment. As explained in the above comments, the major sources of uncertainties are the same for the two methods so that no real improvement in the quality of the results is to be expected. However the results from the two methods will be different and there will be no sure way to discriminate between them. c) As regards the proposed revision of the OBE ground motion assessment, we understand that it would reduce, or possibly suppress, the influence of OBE on design, although final judgement on this point cannot be made before reviewing the draft regulatory guide on piping analyses which should soon be issued for public comment. Our opinion is that this is positive, as we think that the SSE - not the OBE- should be the design-controlling event for safety purposes. Yours sincerely, Remy CARL Executive V" e-President

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 1993

                                                            *93 MA 17 A8 :4 9 rV
                                                                       \I !Cf Mr. Ryo Ikegame Chairman Nuclear Power Development Council The Federation of Electric Power Companies Keidanren Kaikan 1-9-4, Ote-Machi, Chiyoda-Ku Tokyo, Japan

Dear Mr. Ikegame:

On behalf of the Commission, I would like to thank you for your letter of February 4, 1993, concerning the NRC's proposed change to 10 CFR Part 100. Because of the importance of this matter, the Convnission invited convnent on the proposed rule change from various international organizations. We are pleased that the Japanese Federation of Electric Power Companies (JFEPC) will be providing additional convnents on this issue in a separate letter to the Secretary of the Commission. Let me assure you that your convnents will be taken into consideration during the rulemaking process and will be addressed in the final rulemaking documents. Sincerely, Ivan Selin

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THE FEDERATION OF ELECTRIC POWER COMPANIES KEIDANREN KAIKAN 1*8*4,0Tl:*MACHI, CHIYODA*KU, TOKYO,JAPAN Honorable Chairman Ivan Selin US Nuclear Regulatory Commission Washington DC 20555 USA February 4, 1993

Dear Dr. Selin :

As a chairman of the FEPC's(Japanese Federation of Electric Power Companies) Nuclear Power Development Council, I have been paying attention to the importance of the 10 CFR 100 rule change proposed by US Nuclear Regulatory Commission. FEPC represents all nine electric power companies in Japan and is authorized to speak for the utilities on this issue. These nine companies are; The Hokkaido Electric Power Co. ; The Tohoku Electric Power Co. ; The Tokyo Electric Power Co. ; The Chubu Electric Power Co. ; The Hokuriku Electric Power Co. ; The Kansai Electric Power Co. ; The Chugoku Electric Power Co. ; The Shikoku Electric Power Co.; and The Kyushu Electric Power Co .. J APC (Japan Atomic Power Co) also seconds this initiative. During the course of our preliminary review of the proposed rule FEPC identified a number of specific problems. First, the decoupling of reactor siting from reactor design is contrary to the internationally-accepted practice of considering engineered safeguard factors in making siting decision. Consequently, it provides a disincentive to improvements of the safety of reactor designs. Secondly, the numerical criteria for population density and exclusion area boundary distance are unnecessarily restrictive in that they establish a ceiling on population density wholly unrelated to safety goal or to other risks to which the public is exposed. In addition, the numerical criteria, which were initially established in 1975 in Regulatory Guide 4.7 does not reflect the substantial improvements in reactor safety since 1975.

THE FEDERATION OF ELECTRIC POWER COMPANIES KEIDANREN KAIKAN 1*8*4,0TE*MA.CHI, CHIYODA.-KU, TOKYO. JAPAN The proposed rule also requests the use of both the deterministic and the probabilistic seismic analysis approach in determining SSE to allow more informed judgment. Although this approach itself does not seem to be wrong, it is not yet clear if probabilistic seismic analysis methodology and its application in nuclear reactor regulation are matured enough or not. Although it is true that nuclear safety regulation of an individual country remains national responsibility, it is also true that United States had established the basis of LWR safety regulation in the world and will continue to be very influential in the arena of international safety standards. As a chairman of FEPC's Council of Nuclear Power Development, I intend to submit in a separate letter to the Secretary of the Commission comments to this important issue on behalf of the Japanese utility companies and would like to ask you to consider this issue carefully. R~&~ Chairman, Nuclear Power Development Council FEPC cc: Commissioner Curtiss Commissioner de Planque Commissioner Remick Commissioner Rogers

THE FEDERATION OF ELECTRIC POWER COMPANIES KEIDANREN KAIKAN 1*9-4,0TE*MACHI, CHIYOOA*KU, TOKYO. JAPAN Secretary 1

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• U.S. Nuclear Regulatory Commission I ,1 i~l;Cf,i ,t" * , 'v ! f Washington D.C. 20555 USA Attn: Docketing and Service Branch Re: Proposed Rule on Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants March 15, 1993

Dear Mr. Chilk:

The undersigned Chairman of the Nuclear Power Development Council of The Federation of Electric Power Companies of Japan(FEPC), representing the electric power companies of Japan, hereby submits the enclosed comments on the Nuclear Regulatory Commission's proposed rule, "Reactor Site Criteria; Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants", to amend the reactor siting criteria in 10 C.F.R. Part 100. The nine companies under the FEPC are: The Hokkaido Electric Power Co.; The Tohoku Electric Power Co.; The Tokyo Electric Power Co.; The Chubu Electric Power Co.; The Hokuriku Electric Power Co.; The Kansai Electric Power Co.; The Chugoku Electric Power Co.; The Shikoku Electric Power Co.; and The Kyushu Electric Power Co. The Japan Atomic Power Co.(JAPC) joins with the FEPC in submitting the enclosed comments. AY 11 ----. Acknowledged by card ..._..._.__,.. .. er.,,vw

THE FEDERATION OF ELECTRIC POWER COMPANIES KEIDANREN KAIKAN 1*9-4.0TE*MACHL CHIYODA*KU. TOKYO. JAPAN I sincerely hope that the Commission will give serious consideration to the comments of the FEPC in reaching its decision on this most important matter. Sincerely, f,'-.Pt~~ Ry~Ik;/ame, Chairman Nuclear Power Development Council Federation of Electric Power Companies of Japan

Enclosure:

FEPC Comments on Proposed Revisions to NRC Siting Regulations cc: Chairman Ivan Selin Commissioner Kenneth C. Rogers Commissioner James R. Curtiss Commissioner Forrest J. Remick Commissioner E. Gail de Planque

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ENCLOSURE FEPC Comments on Proposed Revisions to NRC Siting Regulations March 15, 1993 (1) The Federation of Electric Power companies of Japan (FEPC) respectfully asks the Nuclear Regulatory Commission to give serious consideration to withdrawing the proposed rule for the following reasons: a) Although it is true that nuclear safety regulation within a particular country remains the national responsibility of that country, it is also true that many countries made reference to the US rule when establishing their rules for LWR safety regulation and the US will continue to be very influential in the arena of international safety standards. The proposed revisions, if adopted, will seriously impact the U.S. nuclear industry, as well as the nuclear industry in other countries. In the earliest days of nuclear reactor siting, the exclusion area was set in relation to core thermal power. Later, however, with the incorporation of engineered safeguards into the design, U.S. siting standards were revised to take these design features into consideration. Many countries with commercial nuclear power plants adopted the U.S. approach. We are confident that this siting approach, together with the other codes, standards and practices to ensure safety, has been sufficient to ensure adequate protection of the health and safety of the public from any undue risk that may arise from the operation of nuclear power plants. By setting certain predetermined numbers for population density and exclusion area, the proposed revisions, if adopted, would reverse this history of ensuring safety through the incorporation of safety technology into the design and would unnecessarily create confusion among the countries using nuclear power. b) The proposed rule decouples siting froin reactor design and, as a result, elevates favorable demographic characteristics over favorable plant design to ensure safety in the selection of the site, thus, effectively cutting the incentive to the nuclear industry for further improvement of reactor safety through design.

c) We understand the underlying philosophy of the Commission's Safety Goals to be the limitation of risk to the public from nuclear power operation by confirming this is not a significant addition to the societal risks. Through application of the Safety Goals, the Commission seeks to improve current practices to provide a better means for testing the adequacy of and need for current and proposed regulatory requirements. However, the proposed rule, by setting limits on population density, would establish additional safety requirements which are beyond the Safety Goals. Thus, the proposed revisions, which are not needed to ensure adequate protection of the public health and safety, would create an inconsistency in the Commission's regulatory philosophy. d) The specific numbers for the population density limits and exclusion area size are taken from revision 1 of Regulatory Guide 4.7, which was issued in 1975. We believe these numbers should be updated in order to take into account the improvements in reactor safety which have been made since 1975. Since specific numbers are subject to change, we believe it is better not to specify this number in a rule. It is better to specify them in the Regulatory Guide. e) The proposed rule, if adopted, would require the use of both deterministic and probabilistic seismic analyses to determine the SSE and allow the use of more informed judgment. The Japanese nuclear industry understands that it is still premature to use the probabilistic approach in a nuclear regulation. Thus the FEPC suggests it will not be necessary to change the rule at this time . (2) Even if there should be an absolute necessity to adopt the proposed rule in conjunction with early site permit reviews, an arrangement should be made so that an Applicant for an early site permit may use either the existing 10 C.F.R. Part 100 rule or the revised rule.

NORTH DAKOTA" GEOLOGICAL SURVEY

                                                              .U::,NHC 600 E. Boulevard Avenue

• Bismarck, North Dakota 58505-0840 Phone (701) 224-4109 FAX (701) 224-3682 John P. Bluemle

                                                    *93 MAR 12 A10 :48 INDUSTRIAL COMMISSION State Geologist                                                                      Edward T . Schafer - Governor, Chairman William A. McClellan                                'J~  !l': ,Jr ':>t. t..tt t 1 ,r-  Heidi Heitkamp     - Attorney General Asst. State Geologist                             [,()Ch[ i !NG .', ~[ 1 v1r f Sarah Vogel              - Commissioner of Agriculture
                                                             ~K NL.Ii January 19, 1993 Dr. Andrew J. Murphy Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Dr. Murphy:

Thank you for giving us the opportunity to review and comment on the revised "Geologic and Seismic Siting Criteria for Nuclear Power Plants." Mr. Thomas Heck, geologist with the North Dakota Geological Survey, has reviewed the seven documents you sent relating these and other issues and we have no comments on the proposed revised criteria. Sincerely, C\D~fJ.~ J;p;;~ ~uemle State Geologist MAY 111~_3_.: Acknowledged by card"....--""'"'" ..........

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                                               *93 MAR 12 A1 0 :48                         George V. Voinovich

• Governor Frances S. Buchholzer
• Director February 16, 1993 Mr. Andrew J . Murphy, Chief Structural and Seismic Engineering Branch Division of Engineering Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Murphy:

Thomas M. Berg, Chief of the Ohio Division of Geological Survey, has asked me to respond to your letter and enclosures regarding a revision to "Geologic and Seismic Siting Criteria For Nuclear Power Plants." We have no critical comments on these documents; however, our technical expertise is limited in some areas. We appreciate the opportunity to review these and other NRC documents. Sincerely,

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Michae l C. Hansen, Ph.D . Senior Geologist Division of Geological Survey MCH:ss MAY t 11993 Acknowledged by card .................................. DNA 0001 Fountain Square

• Columbus, Ohio 43224-1387

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