ML20147J312
| ML20147J312 | |
| Person / Time | |
|---|---|
| Issue date: | 04/22/1997 |
| From: | Ader C, Murphy A NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | NRC |
| Shared Package | |
| ML20013B555 | List: |
| References | |
| FRN-61FR65157, RULE-PR-100, RULE-PR-21, RULE-PR-50, RULE-PR-52, RULE-PR-54 AD93-2-001, AD93-2-1, NUDOCS 9705010238 | |
| Download: ML20147J312 (5) | |
Text
{{#Wiki_filter:1 i j on =rr oy [ k UNITED STATES
<g _I,j NUCLEAR REGULATORY COMMISSION
,p '# WASHINGTON, D.C. 20555 4001
% +
8 April 22, 1997 AD93-2 CF-f)R MEMORANDUM TO: NUDOCS FROM: Andrew J. Murphy, Chief Cr /< 9/up' / Structural & Geological Engineering BrancK Division of Engineering Technology Office of Nuclear Regulatory Research ,
/ / / --
r".arles E. Ader, Chief /2/ , Accident Evaluation Branch bO /" R- - Division of Systems Technology Office of Nuclear Regulatory Research
SUBJECT:
REGULATORY HISTORY - REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR j NUCLEAR POWER PLANTS (10 CFR PARTS 21,50,52,54, AND 100), FINAL RULE Attached is a table that identifies pertinent documentation associated w th the subject rulemaking. Also attached are the documents cited in the table. The documents are listed in two categories, those marked "PDR" that can be made available t: 1he public (items 1 through
- 34) and those marked "CF" that should not be made available to tno public (items 3fthrough 48). Within each category the documents are ordered by date. The designator "AD93-2" has been placed in the upper right-hand corner of each document.
The revision to the regulations was published December 11,1996, in Vol. 61, No. 239 of the Federal Reaister. pages 65157 through 65177 as " Reactor Site Criteria including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants,"(61 FR 65157). The rulemaking effects 10 CFR Parts 21,50,52,54, and 100. Attachments: Table Documents (48)
\
rD\ \ c l aswase na22 22 62ra6 sis 7 gon , LmpusM g,39u e; lipippip,Illpy85 c s , m/ e\ a ncs. n' ann k 'Ma VDi' f , N m I g
+
I l REGULATORY HISTORY Final Rule Revision of 10 CFR Parts 21,50,52,54, and 100 Reactor Site Criteria including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants. (Federal Reaister. Vol. 61, No. 239, pp 65157 - 65177) No. Date Description 1 4/2/93 SECY-93-087, Policy, Technical, and Licensing issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs 2 7/21/93 Memorandum from Samuel J. Chilk to James M. Taylor,
Subject:
SECY 087, Policy, Technical, and Licensing issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs 3 10/17/94 Federal Reaister Notice 59 FR 522%, " 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." 4 2/8/95 Federal Reaister Notice 60 FR 7467, " Reactor Site Criteria including , Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and l Proposed Denial of Petition for Rulemaking From Free iEnvironment, Inc., et al." (extension of public comment period) l 5 2/28/95 Federal Reaister Notice 60 FR 10810, " 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." (extension of public comment period) 6 2/28/95 Federal Reaister Notice 60 FR 10880, " Draft Regulatory Guides and Standard Review Plan Sections; Issuance, Availability" (for public comment) 7 undated Index of commentors on proposed rule and draft regulatory guidance documents. Note: cooies of all comments are in the NRC Public Document Room (PDR). 8 4/3/96 VUGRAPHS, " Draft Final Rule, Revisions of 10 CFR Psrts 50,100," by Leonard Soffer, Andrew J. Murphy, and Nilesh C. Chokshi 9 4/3/96 VUGRAPHS," Staff Presentation on the Part 50 and Part 100 Rulemaking, Sensitivity Analyses," by Barry Zalcman 10 4/11/96 VUGRAPHS, " Draft Final Rule, Revisions of 10 CFR Parts 50,100," by Leonard Soffer and Andrew J. Murphy 11 4/11/96 VUGRAPHS," Staff Presentation on the Part 50 and Part 100 Rulemaking, DBA Dose Calculations, Sensitivity Scaling Analyses," by Barry Zaleman 12 4/11/96 VUGRAPHS, " Revision of 10 CFR Parts 50,52 and 100, by M.W. Gmyrek and R.L. Andersen, Nuclear Energy institute
2 13 4/22/96 Letter from T.S. Kress to The Honorable Shirley Ann Jackson,
Subject:
Proposed Revisions to 10 CFR Parts 50 and 100 and Proposed Regulatory Guides Relating to Reactor Site Criteria 14 5/24/96 SECY-96-118, Amendments to 10 CFR Parts 50,52, and 100, and issuance of a New Appendix S to Part 50 (Attachments 1-9 only, Attachments 10-17 are not available to the public) 15 6/12/96 VUGRAPHS, Commission Briefing, " Draft Final Rule, Revision of 10 CFR Parts 50,100" by Themis Speis, Andrew J. Murphy, and Leonard Soffer 16 7/2/96 Memorandum from John C. Hoyle to James M. Taylor,
Subject:
Staff Requirements - Briefing on Part 100 Final Rule on Reactor Site Criteria (SECY-96-118),3:00 P.M., WEDNESDAY, June 12,1996, Commissioners' Conference Room, One White Flint North, Rockville, Maryland (Open to Public Attendance) 17 7/10/96 Memorandum from James M. Taylor to Chairman Jackson, Commissioner Rogers, and Commissioner Dicus,
Subject:
Response to Staff Requirements Memorandum M960612 - Briefing on Part 100 Final Rule on Reactor Site Criteria 18 10/11/96 Memorandum from John C. Hoyle to James M. Taylor and John F. Cordes,
Subject:
Staff Requirements - Affirmation Session,11:00 A.M., Wednesday, October 2.1996, Commissioners' Conference Room, One white Flint North, Rockville, Maryland (Open to Public Attendance) I 19 12/3/96 Letter to The Honorable Lauch Faircloth from Dennis K. Rathbun (provides Public Announcement and FRN) 20 12/3/96 Letter to The Honorable Dan Schaefer from Dennis K. Rathbun (provides Public Announcement and FRN) 21 12/3/96 Letter to The Honorable Newt Gingrich from Dennis K. Rathbun (provides Final Rule and Regulatory Analysis) 22 12/3/96 Letter to The Honorable Al Gore from Dennis K. Rathbun (provides Final Rule and Regulatory Analysis) 23 12/3/96 Letter to The Honorable Robert P. Murphy from Dennis K. Rathbun (provides Final Rule and Regulatory Analysis) 24 12/11/96 Federal Reaister Notice 61 FR 65157, " Reactor Site Criteria including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants" 25 undated Resolution of Public Comments on the Proposed Seismic and Earthquake Engineering Critoria for Nuclear Power Plants (Section 100.23, Geologic and Seismic Siting Factors to 10 CFR Part 100, and Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants to 10 CFR Part 50, October 17,1994 Publication)
3 26 4/1/97 Federal Reaister Notice 62 FR 15547, " Regulatory Guides and Standard Review Plan Sections; lssuance, Availability"(final guides and SRP sections) 27 3/97 Regulatory Guide 1.165, " Identification and Characterization of Seismic Sources and Determination of Safe Shutr'own Earthquake Ground Motion" 28 3/97 Regulatory Guide 1.12, Revision 2, " Nuclear Power Plant Instrumentation for Earthquakes." 29 3/97 Regulatory Guide 1.166, " Pre-Earthquake Planning and immediate Nuclear Power Plant Operator Postearthquake Actions." 30 3/97 Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event." 31 3/97 Standard Review Plan Section 2.5.1, Revision 3, " Basic Geologic and Seismic Information" 32 3/97 Standard Review Plan Section 2.5.2, Revision 3, " Vibratory Ground Motion." l 33 3/97 Standard Review Plan Section 2.5.3, Revision 3, " Surface Faulting" 34 undated Resolution of Public Comments on Draft Regulatory Guides and Standard Review Plan Sections Pertaining to the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants 35 3/6/96 Memorandum from Themis P. Speis to John T. Larkins,
Subject:
Revisions ) of 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR. Part 50, l New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides l and Standard Review Plan Sections I i 36 4/1/96 Memorandum from Michael T. Lesar to Andrew J. Murphy,
Subject:
Review of Final Rule on Reactor Site Criteria for Ni &n 6wer Plants ) 37 4/1/96 VUGRAPHS, " Briefing to the Chairman, Draft Final Rule Revision of 10 CFR Parts 50,100" 38 4/9/96 Memorandum from Shirley Ann Jackson to James M. Taylor,
Subject:
Proposed Revisions to 10 CFR Part 100 and Part 50 (SECY-94-194) 39 4/30/96 Memorandum from James M. Taylor to Chairman Jackson,
Subject:
Proposed Revisions to 10 CFR Part 100 and Part 50 (SECY-94-194)- Response to Questions 40 undated Explanatory note as a foreword to the Memorandum from David L. Morrison (5/3/96 to Edward L. Jordan,
Subject:
Revisions of 10 CFR Part 100, Reactor Site approx) Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections 41 5/9/96 VUGRAPHS, " Draft Final Rule, Revision of 10 CFR Parts 50,100," by Leonard Soffer and Nilesh C. Chokshi
4 l-f 4 42 5/20/96 Memorandum from Edward L. Jordan to James M. Taylor,
Subject:
Expedited CRGR Review of the Proposed Final Revisions to 10 CFR 100, i 3
" Reactor Site Criteria" ,
43 5/22/96 Memorandum from David L. Meyer to David L. Morrison,
Subject:
Office i Concurrence on Final Rule Entitled " Reactor Site Criteria including Seismic
- and Earthquake Engineering Criteria for Nuclear Power Plants" (10 CFR i Parts 50,52, and 100) 1
{ 44 5/24/96 SECY-96-118, Amendments to 10 CFR Parts 50,52, and 100, and i Ismoce of a New Appendix S to Part 50 (includes all Attachments,1-17) 45 6/4/96 Memorandum from Edward L. Jordan to James M. Taylor,
Subject:
Minutes j of CRGR Meetings Number 285 and 286 46 8/15/96 Memorandum from David L. Morrison to Edward L. Jordan,
Subject:
3 Revisions of 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR ! j Part 50, New Appendix S to Part 50 (Final Rules) and Associated ' j Regulatory Guides and Standard Review Plan Sections ) 47 11/14/96 Memorandum from Michael T. Lesar to Roger Kenneally,
Subject:
Review j of Final Rule Entitled Reactor Site Criteria i e 48 undated Draft Memorandum from James M. Taylor to John C. Hoyle,
Subject:
1 11/19/96 Response to Staff Requirements Memorandum M961002, Staff j j approx) Requirements - Affirmation Session,11:00 A.M., Wednesday, October 2. ! l' 1996, Commissioners' Conference Room, One white Flint North, Rockville, Maryland (Open to Public Attendance), item ll (SECY-96-118 - 3 Amendments to 10 CFR Parts So,52, and 100, and issuance of a New i Appendix S to Part 50) { I l i . l ! o
- _ _ _' __ _ ____ _s __, _ . - _ _ _ _
April 22, 1997 AD93-2
-GF-MEMORANDUM TO: NUDOCS FROM: Andrew J. Murphy, Chief Structural & Geological Engineering Branch Division of Engineering Technology Office of Nuclear Regulatory Research Charles E. Ader, Chief Accident Evaluation Branch Division of Systems Technology Office of Nuclear Regulatory Research
SUBJECT:
REGULATORY HISTORY - REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS (10 CFR PARTS 21,50,52,54, AND 100), ! FINAL RULE I Attached is a table that identifies pertinent documentation associated with the subject l rulemaking. Also attached are the documents cited in the table. The documents are listed in [ two categories, those marked "PDR" that can be made avai!able to the public (items 1 through ;
- 34) and those marked "CF" that should not be made available to the public (items 3fthrough ,
48). Within each category the documents are ordered by date. The designator "AD93-2" has i been placed in the upper right-hand corner of each document. The revision to the regulations was published December 11,1996, in Vol. 61, No. 239 of the . Federal Reaister. pages 65157 through 65177 as " Reactor Site Criteria including Seismic and i Earthquake Engineering Criteria for Nuclear Power Plants," (61 FR 65157). The rulemaking effects 10 CFR Parts 21,50,52,54, and 100. , Attachments: Table Documents (48) Distribution: SGEB r/f RES-2D-1,' RES-2D DOCUMENT NAME: g:\rmk\ hist-finlmem To receive a copy of this document. Indicate in the boa: 'C' = Copy without attachment / enclosure 'E' = Copy with attachment / enclosure "N" = No copy 0FFICE SGEB/RES lE SGEB/RES, / DST /RES lL SGEB/RES l l l NAME RKenneally n d NChokshiW CAder co9 AMurphy 4pau DATE 4/tt /97 V1'I' 4/ (W97 4/ ti /97 4/ u /97 j OFFICIAL RECORD COPY i l y lD"V/ VC)Ju f
- i
oq [,e cer k UNITED STATES j g NUCLEAR REGULATORY COMMISSION
't WASHINGTON, D.C. 20666-0001
\ . . . . . ,d' April 22, 1997 AD93-2 CF MEMORANDUM TO: NUDOCS FROM: Andrew J. Murphy, Chief G[ MW/
Structural & Geological Engineering BrancW Division of Engineering Technology Office of Nuclear Regulatory Research
/ / '
/
Charles E. Ader, Chief // / , Accident Evaluation Branch L d /~6 Division of Systems Technology Office of Nuclear Regulatory Research
SUBJECT:
REGULATORY HISTORY - REACTOR SITE CRITERIA INCLUDING SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS (10 CFR PARTS 21,50,52,54, AND 100), FINAL RULE t Attached is a table that identifies pertinel ' documentation associated with the subject rulemaking. Also attached are the documerits cited in the table. The documents are listed in two categories, those marked "PDR" that can te made available to the public (items 1 through
- 34) and those marked "CF" that should not be made available to the public (items 3fthrough 48). Within each category the documents are ordered by date. The designator "AD93-2" has been placed in the upper right-hand comer of each document.
l The revision to the regulations was published December 11,1996, in Vol. 61, No. 239 of the Endtral Reoister. pages 65157 through 65177 as " Reactor Site Criteria Including Seismic and Eartaquake Engineering Criteria for Nuclear Power Plants," (61 FR 65157). The rulemaking effec's 10 CFR Parts 21, 50, 52, 54, and 100. Attachments: Table Documents (48) , l l l l I 9 705 0/e d h% S PP
y.-. - - - -.-. -.-.-. ..- - _.-._.-.-. - . . . - _ . - . - . - _ . . . - - - - . 1 ['.. t
- REGULATORY HISTORY
] Final Rule j Revision of 10 CFR Parts 21,50,52,54, and 100 t j Reactor Site Criteria including Seismic and Earthquake Engi sering Criteria i for Nuclear Power Plants. t l (Federal Reaister. Vol. 61, No. 239, pp 65157 - 65177) i 1 i No. Date Description 2
- 1 4/2/93 SECY-93-087, Policy, Technical, and Licensing issues Pertaining to Evolutionary and Advanced Light-Water Reactor (ALWR) Designs i 2 7/21/93 Memorandum from Samuel J. Chilk to James M. Taylor,
Subject:
SECY 1 087, Policy. Technical, and Licensing issues Pertaining to Evolutionary and l Advanceyght-Water Reactor (ALWR) Designs l 3 10/17/94 Federal Reaister Notice 59 FR 52255, " Reactor Site Criteria Including i Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and i Proposed Denial of Petition for Rulemaking From Free Environment, Inc., et j al." l 4 2/8/95 Federal Reaister Notice 60 FR 7467, " Reactor Site Criteria including i Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and , Proposed Denial of Petition for Rulemaking From Free Environment, Inc., et j al." (extension of public comment period) j 5 2/28/95 Federal Reaister Notice 60 FR 10810, " Reactor Site Criteria Including , i Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and i j Proposed Denial of Petition for Rulemaking From Free Environment, Inc., et l ! al." (extension of public comment period) ' , I l 6 2/28/95 Federal Reaister Notice 60 FR 10880, " Draft Regulatory Guides and Standard Review Plan Sections; issuance, Availability" (for public comment) i 7 undated index of commentors on proposed rule and draft regulatory guidance 2 documents. Note: copies of all comments are in the NRC Public Document Room (PDR). 8 4/3/96 VUGRAPHS, " Draft Final Rule, Revisions of 10 CFR Parts 50,100," by
- Leonard Soffer, Andrew J. Murphy, and Nilesh C. Chokshi
- 9, 4/3/96 VUGRAPHS," Staff Presentation on the Part 50 and Part 100 Rulemaking, j Sensitivity Analyses," by Barry Zaleman i 10 4/11/96 VUGRAPHS, " Draft Final Rule, Revisions of 10 CFR Parts 50,100," by 2
Leonard Soffer and Andrew J. Murphy
- 11 4/11/96 VUGRAPHS," Staff Presentation on the Part 50 and Part 100 Rulemaking, j DBA Dose Calculations, Sensitivity Scaling Analyses," by Barry Zaleman f 12 4/11/96 VUGRAPHS, " Revision of 10 CFR Parts 50,52 and 100, by M.W. Gmyrek
; and R.L. Andersen, Nuclear Energy Institute i
q .. 2 13 4/22/96 Letter from T.S. Kress to The Honorable Shiriey Ann Jackson,
Subject:
4 Proposed Revisions to 10 CFR Parts 50 and 100 and Proposed Regulatory : Guides Relating to Reactor Site Criteria l 14 5/24/96 SECY-96-118, Amendments to 10 CFR Parts 50,52, and 100, and issuance of a New Appendix S to Part 50 (Attachments 1-9 only, Attachments 10-17 are not available to the public) 15 6/12/96 VUGRAPHS, Commission Briefing, " Draft Final Rule, Revision of 10 CFR : Parts 50,100" by Themis Spels, Andrew J. Murphy, and Leonard Soffer , 16 7/2/96 Memorandum from John C. Hoyle to James M. Taylor,
Subject:
Staff Requirements - Briefing on Part 100 Final Rule on Reactor Site Criteria (SECY-96-118),3:00 P.M., WEDNESDAY, June 12,1996, Commissioners' Conference Room, One White Flint North, Rockville, Maryland (Open to l 1 l Public Attendance) 17 7/10/% Memorandum from James M. Taylor to Chairman Jackson, Commissioner Rogers, and Commissioner Dicus,
Subject:
Response to Staff , Requirements Memorandum M%0612 - Briefing on Part 100 Final Rule on Reactor Site Criteria 18 10/11/96 Memorandum from John C. Hoyle to James M. Taylor and John F. Cordes, i
Subject:
Staff Requirements - Affirmation Session,11:00 A.M., Wednesday, October 2.1996, Commissioners' Conference Room, One white Flint North, l Rockville, Maryland (Open to Public Attendance) 19 12/3/96 Letter to The Honorable Lauch Faircloth from Dennis K. Rathbun (provides l Public Announcement and FRN) ] 20 12/3/96 Letter to The Honorable Dan Schaefer from Dennis K. Rathbun (provides I Public Announcement and FRN) l 21 12/3/96 Letter to The Honorable Newt Gingrich from Dennis K. Rathbun (provides l Final Rule and Regulatory Analysis) l 22 12/3/96 Letter to The Honorable Al Gore from Dennis K. Rathbun (provides Final Rule and Regulatory Analysis) 23 12/3/96 Letter to The Honorable Robert P. Murphy from Dennis K. Rathbun (provides Final Rule and Regulatory Analysis) 24 12/11/96 Federal Reaister Notice 61 FR 65157, " Reactor Site Criteria Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants" 25 undated Resolution of Public Comments on the Proposed Seismic and. Earthquake 1 Engineering Criteria for Nuclear Power Plants (Section 100.23, Geologic j and Seismic Siting Factors to 10 CFR Part 100, and Appendix S, ! Earthquake Engineering Criteria for Nuclear Power Plants to 10 CFR Part 50, October 17,1994 Publication)
l j i ! 3 : i . 26 4/1/97 Federal Reaister Notice 62 FR 15547, " Regulatory Guides and Standard j Review Plan Sections; Issuance, Availability" (final guides and SRP ; I' sections) 27 3/97 Regulatory Guide 1.165, " identification and Characterization of Seismic l Sources and Determination of Safe Shutdown Earthquake Ground Motion" 28 3/97 Regulatory Guide 1.12, Revision 2, " Nuclear Power Plant instrumentation for Earthquakes." ; 29 3/97 Regulatory Guide 1.166, " Pre-Eadhquake Planning and immediate Nuclear Power Plant Operator Postearthquake Actions." ! 30 3/97 Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event." 31 3/97 Standard Review Plan Section 2.5.1, Revision 3, " Basic Geologic and ) Seismic Information" l l 32 3/97 Standard Review Plan Section 2.5.2, Revision 3, " Vibratory Ground Motion." 33 3/97 Standard Review Plan Section 2.5.3, Revision 3, " Surface Fautting" 34 undated Resolution of Public Comments on Draft Regulatory Guides and Standard Review Plan Sections Pertaining to the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants 35 3/6/96 Memorandum from Themis P. Speis to John T. Larkins,
Subject:
Revisions of 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections 36 4/1/91 Memorandum from Michael T. Lesar to Andrew J. Murphy,
Subject:
Review of Final Rule on Reactor Site Criteria for Nuclear Power Plants 37 4/1/96 VUGRAPHS, " Briefing to the Chairman, Draft Final Rule Revision of 10 CFR Parts 50,100" 38 4/9/96 Memorandum from Shirley Ann Jackson to James M. Taylor,
Subject:
Proposed Revisions to 10 CFR Part 100 and Part 50 (SECY-94-194) 39 4/30/96 Memorandum from James M. Taylor to Chairman Jackson,
Subject:
Proposed Revisions to 10 CFR Part 100 and Part 50 (SECY-94-194)- Response to Questions 40 undated Explanatory note as a foreword to the Memorandum from David L. Morrison (5/3/96 to Edward L. Jordan,
Subject:
Revisions of 10 CFR Part 100, Reactor Site approx) Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections 41 5/9/96 VUGRAPHS, " Draft Final Rule, Revision of 10 CFR Parts 50,100," by Leonard Soffer and Nilesh C. Chokshi
,. l 4
42 5/20/96 Memorandum from Edward L. Jordan to James M. Taylor,
Subject:
Expedited CRGR Review of the Proposed Final Revisions to 10 CFR 100,
" Reactor Site Criteria" 43 5/22/96 Memorandum from David L. Meyer to David L. Morrison,
Subject:
Office Concurrence on Final Rule Entitled " Reactor Site Criteria Including Seistnic and Earthquake Engineering Criteria for Nuclear Power Plants" (10 CFR Parts 50, 52, and 100) 44 5/24/96 SECY-96-118, Amendments to 10 CFR Parts 50, 52, and 100, and issuance of a New Appendix S to Part 50 (includes all Attachments,1-17) 45 6/4/96 Memorandum from Edward L Jordan to James M. Taylor,
Subject:
Minutes of CRGR Meetings Number 285 and 286 46 8/15/96 Memorandum from David L. Morrison to Edward L. Jordan,
Subject:
Revisions of 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections 47 11/14/96 Memorandum from Michael T. Lesar to Roger Kenneatly,
Subject:
Review of Final Rule Entitled Reactor Site Criteria 48 undated Draft Memorandum from James M. Taylor to John C. Hoyle,
Subject:
11/19/96 Response to Staff Requirements Memorandum M961002, Sta# approx) Requirements - Affirmation Session,11:00 A.M., Wednesday, October 2. 1996, Cornmissioners' Conference Room, One white Flint North, Rockville, Maryland (Open to Public Attendance), item ll (SECY-96-118 - Amendments to 10 CFR Parts 50,52, and 100, and issuance of a New Appendix S to Part 50) i
Ap93 T - m<cg PDS i 9j
% ..... / '
POLICY ISSUE April 2, 1993 SECY-93-087 ; f98: The Cossaissioners fBQH: James M. Taylor Executive Director for Operations
SUBJECT:
POLICY, TECHNICAL, AND LICENSING ISSUES PERTAINING TO EVOLU-TIONARY AND APVANCED LIGHT-WATER REACTOR (ALWR) DESIGNS PURPOSE: To present the Cossaission with recomunended positions pertaining to evolution-ary and passive light-water reactor LWR and to request that the Cosumission ap(prov)e the underlined staff positionsdesign certi presented in this paper. SW95RY: In Enclosure 1, the Nuclear Regulatory Commission staff discusses 42 technical and policy issues pertaining to either evolutionary LWRs, passive LWRs, or both. The staff previously identified these issues in the draft Commission papers, " Issues Pertaining to Evolutionary and Passive Light-Water Reactors and Their Relationship to Current Regulatory Requirements," dated February 20, 1992, and " Design Certification Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light-Water Reactor Designs," dated June 25, 1992. After consider!ng the Advisory Committee on Reactor Safeguards (ACRS), industry, and vandor comuments, the staff has reached a f' nal position on many of the issues and the staff has underlined the positions for uhich it is requesting the Casumission's approval. The staff also discusses other issues, CONTACTS: 4 N: E BE MADE N M M N IN 3 WORKING DAYS FROM THE DATE Thomas G. Hiltz M M S N ER 1 504-1105 I [ "
~
l . I The Commissioners f
- which it concludes may be of interest to the Commission. For these issues, i the staff will update the Commission, if appropriate, after reaching a final staff position or as warranted by new, substantive inforamtion on the issues.
! BACKGROUND: I The staff has forwarded several policy papers' to the Commission proposing resolutions for policy matters and major technical issues concerning both ) evolutionary and advanced LWR designs. Two draft Commission papers, cited in the above summary, were released to the public after they were forwarded to the Commission. Those papers supported dialogue between the NRC staff, the 1 ACRS, the Electric Power Research Institute (EPRI), the vendors, and other j industry representatives which was focused on resolving the policy and technical issues. - ! The staff has considered the comments received and reached a final position on i many of the issues listed in the draft Commission sapers. The staff has j finalized its position on 20 issues and requests tiat the Commission approve ! the positius recosusended in this paper. The staff has also deterinined that 8 9 issues discussed in the draft Commission papers are not policy issues and )' ! the staff does not anticipate any future interaction with the Commission ! concerning these issues. For these 9 issues and for the 13 issues which the i staff will discuss its final position in future Commission papers, the staff j discussions are for the Commission's information only. j DISCUSSION: i Enclosure I discusses the staff's positions and the current regulatory i requirement or interpretation, as well as comments received from the ACRS, the , industry, and the vendors reganling 42 technical and policy issues pertaining to evolutionary LWR designs, passive LWR designs, or both. Where appropriate, the staff has included a detailed discussion of the basis for its position on .! each issue. The staff has also underlined the positions for which it is ! requesting the Commission's approval. 4 Enclosure I is divided into three sections. Section I discusses issues previously identified to the Commission in SECY-90-016, " Evolutionary Light-Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements." Section II discusses other evolutionary and passive LWR design issues and Section III discusses issues which are applicable only to passive LWR designs. Preliminary analysis and reconnendations for issues discussed in Sections II and III were previously transmitted to the Commission in draft Commission papers " Issues Pertaining to Evolutionary and Passive
' Enclosure 3 lists the papers that the staff has forwarded to the Commission regarding policy issues identified for evolutionary and passive advanced light-water reactors. The staff references applicable documents throughout this paper.
1 3
The Connissioners ! i Light-Water Reactors and Their Relationship to Current Regulatory Require-ments," dated February 20, 1992, and " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light-Water Reactor Designs," dated June 25, 1992. Enclosure 2 cross-references issues pertaining to evolutionary, and passive LWR designs with the Comunission papers in which the staff has addressed each issue. Enclosure 3 lists Commission papers related to ALWR designs. The staff developed the recommendations in this paper after: (1) reviewing current operating reactor designs, evolutionary designs, and passive ALWR design infonnation which was available; (2) considering operating experience; (3) considering insights from the available results of the probabilistic risk assessments (PRAs) of LWRs and ALWRs; (4) considering the Commission's guidance on issues resolved for the evolutionary ALWRs; (5) completing the draft safety evaluation report for the EPRI Utility Requirements Document (URD) for passive ALWR designs; (6) completing the final safety evaluation report for the EPRI URD for evolutionary ALWR designs; and, (7) considering EPRI, ACRS, and industry comments on these issues. The staff concludes that the positions discussed in Enclosure 1 are funda-mental to the Agency's decisions on the acceptability of the evolutionary and passive LWR designs. As discussed in SECY-gl-262, " Resolution of Selected Technical and Severe Accident Issues for Evolutionary Light-Water Reactor (LWR) Designs," the staff proposes to implement final positions on these matters as approved by the Commission through individual design certifications and generic rulemaking, as appropriate. The staff plans to forward to the Commission and solicit ACRS and industry comments on at least two additional Commission papers which will discuss issues relating to (1) the regulatory treatment of nonsafety systems in passive designs and (2) use of a physically based source ters. CGEMISIONS: The staff requests that the Commission approve the recommended positions for issues pertaining to evolutionary LWR designs. Such approval would enable the staff to proceed with the final design approval and the design certification review of GE Nuclear Energy's (GE) Advanced Boiling Water Reactor and Asea Brown Boveri-Combustion Engineering's (AB8-CE) System 80+ LWR designs.
. . . _ . _ ~ _ _ _ . _ . _ . _ . _ . _ . . _ _ _ _ ._ _ . - . . _ _ _ _ _ . _ . _ _ . _ _ _ . . _ _ _ _ _
i
- The Commissioners !
The staff also requests that the Conaission approve the proposed positions for ! issues pertaining to the passive designs. This will enable the staff to
! proceed more effectively with its review of Westinghouse's AP600 and GE's l Simplified Boiling Water Reactor ALWR designs.
1 COORDINATION: 1 i i The Office of General Counsel (0GC) has reviewed this paper and has no legal objection. 0GC notes that Commission approval would be tentative, subject to a further review in design certification rulemakings, and that comernications j with vendors and EPRI regarding these Commission positions shoulc. state this fact. RECOMMENDATIONS: i j The staff recommends that the Commission f (1) Anorove the positions underlined in Enclosure 1. (2) Holt that the staff is still considering other policy issues and it will seek the Commission's approval of its positions in the future. i f ~ \ 4 wY.A i <ames M. Taylor l ! j W.xecutive Director for Operations 4
Enclosures:
j 1. Policy Issues Analysis j and Recommendations ! 2. ALWR Issue Cross- l ! Reference Matrix '
- 3. Commission Papers I
- Applicable to i ALWRs.
l 1 ! 1 1 i i i -: l 1 1
^
f Commissioners' coments or consent should be provided directly to the Office of the Secretary by COB Monday, April 19, 1993. Cossaission Staff Office comments, if any, should be submitted to the Comissioners NLT Monday, April 12, 1993, with an infor-nation copy to the office of the Secretary. If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be expected. DISTRIBUTION: Commissioners OGC ' OCAA OIG ' OPA OCA OPP EDO ACRS ' SECY \
i TABLE OF CONTENTS i Eass ! I. SECY-90-016 Issues ......................... I A. Use of a Physically Based Source Term . . . . . . . . . . . . . 1 Anticipated Transient Without Scram . . . . . . . . . . . . . . J B. 4
- C. Mid-Loop Operation ...................... 5
- D. Station Blackout ....................... 6
- E. Fire Protection ....................... 7 i F. Intersystem Loss-of-Coolant Accident ............. 9 i G. Hydrogen Control ....................... 10 i H. Core Debris Coolability . . . . . . . . . . . . . . . . . . . . 12 l I. High-Pressure Core Melt Ejection ............... 14 i J. Containment Performance . . . . . . . . . . . . . . . . . . . . 15 1 K. Dedicated Containment Vent Penetration ............ 18 L. Equipment Survivability ................... 19 M. Elimination of Operating-Basis Earthquake .......... 20 j N. Inservice Testing of Pumps and Valves ............ 26
) 3 II. Other Evolutionary and Passive Design Issues . . . . . . . . . . . . 28 A. Industry Codes and Standards ................. 28-1 B. Electrical Distribution . . . . . . . . . . . . . . . . . . . . 28 ) C. Seismic Hazard Curves and Design Parameters . . . . . . . . . . 29 Leak-Before-Break D. ...................... 30 Classification of Main Steamlines in Boiling Water Reactors . . E. 34
- F. Tornado Design Basis ..................... 37
- G. Containment Bypass ...................... 38 4
H. Containment Leak Rate Testing ................ 39 I. Post-Accident Sampling System ................ 40 ! J. Level o f De t a i l . . . . . . . . . . . . . . . . . . . . . . . . 43
- K. Prototyping . . . . . . . . . . . . . . . . . . . . . . . . . . 44
- L. ITAAC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 i
M. Reliability Assurance Program . . . . . . . . . . . . ... . . . 46 i N. Site-Specific ProbaMilstic Risk Assessments and Analysis of j External Events . ...................... 48 j 0. Severe Accident Mitigation Design Alternatives ........ 52 . P. Generic Rulemaking Related to Design Certification ....., 53 i Q. Defense Against Common-Mode Failures in Digital Instrumenta-tion andControl Systems ................... 54 R. Steam Generator Tube Ruptures ................ 58
- S. PRA Beyond Design Certification . . . . . . . . . . . . . . . . 64 1 T. Control Room Annunciator (Alare) Reliability ......... 66
{ III. Issues Limited to Passive Designs . . . . . . . . . . . . . . . . . 68 A. Regulatory Treatment of Nonsafety Systems in Passive Desi 68 B. Definition of Passive Failure . . . . . . . . . . . . . .gns ... 70 i C. SBWR Stabflity ........................ 71 D. Safe Shutdown Requirements .................. 72 E. Control Room Habitability . . . . . . . . . . . . . . . . . . . 73 F. Radionuclide Attenuation ................... 75 G. Simplification of Offsite Emergency Planning ......... 76 H. Role of the Passive Plant Control Room Operator ....... 77
POLICY ISSUES ANALYSIS AND RECOMMENDATIONS This enclosure is divided into three sections. Section I discusses issues previously identified to the Commission in SECY-90-016, " Evolutionary Light-Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements." Section II discusses other evolutionary and passive i LWR design issues and Section III discusses issues which are applicable only to passive LWR designs. Preliminary analysis and recommendations for issues discussed in Sections II and III were previously transmitted to the Commission in draft Commission papers " Issues Pertaining to Evolutionary and Passive Light-Water Reactors and Their Relationship to Current Regulatory Requirements," dated February 20, 1992, and " Design Certification and i Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light- ! Water Reactor Designs," dated June 25, 1992. The staff has reviewed comments from industry and the Advisory Committee on Reactor Safeguards (ACRS) regarding the staff positions discussed in SECY-90-016 and the preliminary staff positions discussed in the two draft
)
Commission papers. The staff has evaluated these comments, continued to dialogue with the industry and the ACRS and, where appropriate, revised our discussions to address industry and ACRS comments. The staff requests that the Commission review the discussions contained in this enclosure and approve the underlined staff positions. I. SECY-90-016 Issues A. Use of a Physically Based Source Term This section provides an overview of the current source term status and discusses source tem policy issues previously identified in SECY-90-016. A , complete source tem discussion pertaining to advanced light-water reactor l (ALWR) design will be presented in the forthcoming Commission paper on source ters. For approximately 30 years, the Nuclear Regulatory Commission (NRC) staff has been using the reactor accident source term guidelines contained in Technical i Information Document (TID) 14844, " Calculation of Distance Factors for Power ! and Test Reactor Sites (March 1962)," to evaluate design-basis-accident (OBA) ' analyses. In SECY-90-016, the staff discussed the methodology for detemining compliance with the siting requirements of 10 CFR Part 100 using TID-14844. The staff noted that the assumptions in the TID-14844 methodology considered the uncertainties associated with accident sequences and equipment performance and were intended to ensure that future plant sites would provide sufficient safety margins. This methodology has remained essentially unchanged and many i of its original assumptions are considered outdated. In SECY-90-016, the staff recommended that the Commission approve the l following approach for evolutionary ALWRs: Enclosure 1
- 1. Ensure that evolutionary designs meet the requirements of 10 CFR Part 100.
- 2. Consider deviations from current methodology used to calculate 10 CFR Part 100 doses on a case-by-case basis using engineering judgement including updated information on source ters and equipment reliability.
- 3. Do not modify current siting practice.
- 4. Continue to interact with the Electric Power Research Institute (EPRI) and the evolutionary ALWR vendors to reach agreement on the appropriate use of updated source term information for severe accident performance considerations.
In its staff requirements memorandum (SRM) of June 26, 1990, the Commission approved the staff's approach in detemining the source term for the evolutionary designs. The Commission also directed the staff to modify regulations, regulatory practices, and the review process, as appropriate, to reflect information resulting from source term research. As a result of this guidance, the NRC staff has developed a new source term based on calculations performed using the source term code package for individual accident sequences selected in NUREG-Il50, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," dated December 1990. The proposed new source term is discussed in NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants" and was issued in July 1992 as a draft report for public comment. NUREG-1465 discusses radionuclide release fractions, release timing, ano chemical forms of fission products that would be released into containment based upon a range of core melt accident scenarios, including failure of the reactor vessel and subsequent molten core-concrete interactions. The new source term document was not intended and does not provide a specific , methodology or implementation guidelines for the staff's licensing revf ew of advanced light-water reactors. Needed information includes quantifiestion of parameters related to the removal or reduction of fission products within containment via the use of engineered safety features such as sprays and filters, and passive processes such as aerosol deposition and plateout. The NRC staff is currently working with Sandia National Laboratory to evaluate fission product removal mechanisms within containment and quantify associated releases to the environment. EPRI has proposed source tems which are based on bounding severe reactor accidents. EPRI provided the staff with technical justification for their new : source tems in correspondence dated October 18, 1990, and February 12, 1991, ! entitled " Licensing Design Basis Source Term Update for the Evolutionary ALWR," and " Passive ALWR Source Tem," respectively. The EPRI-proposed source tems are based on single, enveloping values for bounding severe reactor l
l ! I i accident sequences, using release data obtained from (1) the Severe Fuel Damage Tests at the Power Burst Facility, (2) the Loss of Fluid Test (LOFT) l source term measurements, and 3) data from the THI-2 post accident examination. l l l l For the evolutionary designs, the EPRI source terms address fission product i fuel release magnitude, the fission product release timing, the chemical form 1 ! of iodine, the retention of aerosol in the reactor coolant and the use of the l l suppression pool and containment sprays for removal of aerosol and soluble i ! gases. For the passive designs, EPRI proposes that the source terms also l consider passive mitigation functions and systems such as steam condensation- ! driven aerosol removal, main steam isolation valve leakage control, and
- secondary building fission product leakage control.
The staff compared the draft source terms in NUREG-1465 with EPRI evolutionary and passive LWR source terms specifically focusing on the accident severity l selected, the nature of the release phases, and the timing and magnitude of ) the important nuclides released. The staff concludes that there is, in general, good agreement between the EPRI-proposed source term for the passive plants and the staff's proposed source term. However, the staff noted that the staff values for the magnitude of low volatility radionuclides were about an order of magnitude higher than those proposed by EPRI. These differences appear to be attributed to EPRI's assumption that little or no fission ' products will be released to the containment atmosphere from ex-vessel debris following core-concrete interaction. According to EPRI, this is due to the ability of the passive ALWR to provide ample coolant' to the reactor cavity / ) lower drywell prior to or immediately upon vessel penetration. The NRC staff l model does not make this assumption. j GE Nuclear Energy (GE) demonstrated in its standard safety analysis report l that the Advanced Boiling Water Reactor (ABWR) design will meet the offsite dose reference values set forth in 10 CFR Part 100 using the current TID-14844 source ters. The staff has reviewed the ABWR design and performed an independent analysis of the radiological consequences resulting from a postulated DBA and concluded in the draft final safety evaluation that the ABWR design will meet the dose reference values set forth in 10 CFR Part 100. Asea Brown Bovert-Combustion Engineering, Inc. (ABB-CE) initially proposed the use of the TID-14844 source ters and the existing analyses in the ABB-CE System 80+ standard safety analysis report are based on the TID-14844 source ters. However, ABB-CE is evaluating the possibility of adopting the NUREG-1465 source ters. In its AP600 design, Westinghouse proposed the same accident source ters as roposed by EPRI for passive plants. In its simplified boiling water reactor (SBWR) design, GE proposed the same accident source ters as published in NUREG-1465. The staff continues to interact with the passive plant vendors to resolve source-term related issues. l
l i l ) The staff will continue its approach for review of source terms proposed by EPRI and evolutionary and passive plant vendors (as approved in the June 26, 1990, SRM) without waiting for final agency adoption of NUREG-1465. The staff evaluations of ALWR submittals will utilize the current source term research i insights regarding fission product release into containment. In determining l the effects of removal mechanisms such as sprays, filters, plateout and , aerosol deposition, the staff will utilize engineering judgement and best l estimates for the applicable parameters. ! The staff is considering several source term related design certification issues. The staff previously identified several source tem related policy issues involving control room habitability, radionuclide attenuation, and containment bypass in two draft Commission papers. In addition, several new source term related policy issues appear to be developing which may concern (1) dose assessments for ex-vessel. releases which result from a severe accident scenario (inclusion of release from core-concrete interaction), (2) assessment of safety-related equipment qualification and post-accident sampling capabilities and vital area access prov(3) assessment isions. The of staff's proposed resolution of these issues will be discussed in a separate Commission paper which will discuss source term related issues. B. Anticipated Transient Without Scram As discussed in SECY-90-016, the ATWS Rule (10 CFR 50.62) was promulgated to reduce the probability of an anticipated transient without scram (ATWS) and to enhance mitigation capability if such an event occurred. The staff recommended that the Commission approve its position that diverse scram systems should be provided for evolutionary ALWRs. In addition, the staff indicated that GE would perform a reliability analysis to detemine whether they could justify manual operation of the standby liquid control system (SLCS), in lieu of automatic operation as required by 10 CFR 50.62, in the event of an ATWS. In its SRM of June 26, 1990, the Commission approved the staff's position. However, the Commission directed that the staff should accept an applicant's alternative to the diverse scram system, if the applicant can demonstrate that the consequences of an ATWS are acceptable. The A8WR design includes a number of features that reduce the risks associated with an ATWS event. These features include a diverse scram system with both hydraulic and electric run-in capabilities on the control rods, a SLCS, and a recirculation pump trip capability. In its letter dated October 9,1991, GE indicated that it will automate the SLCS and they subsequently provided this infomation in amendment 20 to the SSAR. In addition, the scram discharge volume has been removed from the ABWR design, eliminating some of the potential ATWS problems associated with the older boiling water reactor (BWR) designs. f
, . _ . . . . , - - _ _ - - . - ,- ~ . - . . - . _
l . . t l The A88-CE System 80+ design includes a control-grade alternate protection system which is separate and diverse from the safety-grade reactor trip , system. This system provides an alternate reactor trip signal and an alternate feedwater actuation signal. 1 The staff has evaluated the GE ABWR and the ABB-CE System 80+ evolutionary LWR designs and concludes that the designs adhere to the Commission's guidance regarding diverse scram systems. l l In its letter of December 6,1991, EPRI stated that it has detennined that automatic actuation of the SLCS was appropriate, and that it was modifying the . requirements document for evolutionary designs to reflect that position. EPRI ' no longer considers this to be a plant optimization subject. > In its requirements documents, EPRI provides design requirements that are
~
l ! consistent with the staff's position on ATWS discussed in SECY-90-016, as i modified by the Commission's SRM of June 26, 1990. In its letter of May 5, t 1992, EPRI indicated that its approach to resolving the ATWS issue, for both ; evolutionary and passive designs, is compliance with the ATWS Rule. EPRI has i not proposed design requirements beyond those required to meet the rule. The passive ALWR vendors have indicated that their designs will comply with the requirements document for passive designs, and the staff is evaluating ; passive designs to ensure compliance with Commission regulations and guidance i regarding ATWS. The staff considers this policy issue resolved. ! C. Mid-Loop Operation t l In SECY-90-016, the staff stated its concern that decay heat removal l capability could be unavailable when a pressurized-water reactor (PWR) is shut down for refueling or maintenance and drained to a reduced reactor coolant system (RCS) or "mid-loop" level. The staff reconwnded that the Commission approve its position that evolutionary PWR vendors must propose design features to ensure high reliability of the shutdown decay heat removal system. In its letter of April 26, 1990, the ACRS recommerAed that the staff consider four additional requirements to resolve this isnue. In its memorandum of April 27, 1990, the staff indicated that it would ensure that the four ACRS l recommendations would be considered during the review of evolutionary PWR designs. In its SRM of June 26, 1990, the Commission approved the staff's position and endorsed consideration of the four additional ACRS requirements for mid-loop operation. However, the staff remained concerned that the vendors and EPRI have not adequately evaluated the overall question regarding the vulnerability .
; of the ALWRs during shutdown and low-power operetion. This issue was !
l discussed with regard to evolutionary designs in a memorandum to the i ! I s 1 i l
1 . 4 . I j l
; i 3
j Commission dated September 5, 1990. The staff requested that ALWR vendors and l i EPRI assess shutdown and low-power risk, identifying design-specific i vulnerabilities and weaknesses and documenting their consideration and ; incorporation of design features that minimize such vulnerabilities. j i In its letter of December 16, 1991 EPRI :uomitted proposed changes to the : requirements document to address this issue for the evolutionary designs. In ' its letter of May 5, 1992, EPRI stated that the Passive Requirements Document ; ! specifies extensive deterministic requirements to address known shutdown risks i j based on industry experience. EPRI also requires probabilistic and ' , operational shutdown risk evaluations and analysis and has submitted a i revision to the requirements documents to address additional requirements j resulting from its review of NUREG-1410 " Loss of Vital AC Power and the j Residual Heat Removal System During Midloop Operations at Vogtle Unit 2 on 3 March 20, 1990," and NUREG 1449, " Shutdown and Low-Power Operation at j Commercial Nuclear Power Plants in the United States." The EPRI requirements document and evolutionary PWR designers have provided i
- features to address the issue of mid-loop operation. The staff has reviewed I l
the features for the AB8-CE System 80+ design and concludes that this design l adequately implements the Commission's guidance on this issue. ! i ! j The staff concludes that passive plants must also have a reliable means of ' 1 maintaining decay heat removal capability during all phases of shutdown ) activities, including refueling and maintenance, and will evaluate the adequacy of designs during its review. The staff does not consider this issue ! to be a policy matter, but rather an element of its normal review. Therefore, j the staff considers this policy issue resolved. ] D. Station Blackout i 'As discussed in SECY-90-016, the Station Blackout Rule (10 CFR 50.63) allows i utilities several design alternatives to ensure that an operating plant can ] safely shut down in the event that all ac power (offsite and on-site) is i unavailable. The staff concluded that the preferred method of demonstrating j compliance with 10 CFR 50.63 is through the installation of A spare (full-j capacity) alternate ac power source of diverse design. This power source j should be consistent with the guidance in RG 1.155, and should be capable of
- powering at least one complete set of normal shutdown loat.s. The staff .
recossended that the Commission approve its position mandating an alternative , ac power source for evolutionary ALWRs. In its SRM of June 26, 1990, the '
- Commission approved the staff's position.
! In addition to other design features to address the issue of station blackout, the EPRI requirements document for evolutionary designs and the evolutionary i ALWR vendors have provided for a large-capacity, alternative ac power source (combustion turbine generator) with the capability to power one complete set of normal safe-shutdown loads. The staff concludes that the EPRI proposal
- that calls for t. " combustion turbine unit" which EPRI has concluded would l " achieve diversity of power sources and maximize the overall reliability of 4
1
the on-site standby ac power supply system" meets the intent of the Commission guidance on this issue. The staff is still in the process of evaluating the ALWR vendor submittals to ensure acceptable implementation of the Commission's . l guidance on this issue, but does not expect any related policy matters to result from its review. Because the passive ALWR designs do not rely on active systems for safe shutdown following an event, EPRI and the passive plant designers have indicated that both safety-related diesel generators and an alternate ac power source should not be required. However, the staff believes that the diesel generators included in passive LWR designs may require some regulatory treatment. The staff is still evaluating this issue for the passive plant designs. The ' staff's proposed resolution of this issue will be discussed in a separate i Commission paper which will discuss the regulatory treatment of nonsafety i systems in passive plant designs. E. Fire Protection 1 As discussed in SECY-90-016, the staff recommended that current NRC guidance 1 to resolve fire protection issues should be enhanced to minimize fire as a i significant contributor to the likelihood of severe accidents for advanced 1 plants. The staff proposed to require that evolutionary ALWR designers must ensure that safe shutdown can be achieved assuming that all equipment in any one fire area will be rendered inoperable by fire and that re-entry into the fire area for repairs and operator actions is not possible. Because of its physical configuration, the control room is excluded from this approach, provided an independent alternative shutdown capability that is physically and electrically independent of the control room is included in the design. Evolutionary ALWR designers must provide fire protection for redundant shutdown systems in the reactor containment building that will ensure, to the extent practicable, that one shutdown division will be free of fire damage. Additionally, evolutionary ALWR designers must ensure that smoke, hot gases, or the fire suppressant will not migrate into other fire areas to the extent that they could adversely affect safe shutdown capabilities, including operator actions. In its letter of April 26, 1990, the ACRS recommended that the staff consider additional matters in its evaluation of the fire protection designs such as redundant train separation. In its response to the ACRS letter dated April 27, 1990, the staff stated that the proposed requirement to consider the effects of smoke, heat, and fire suppressant migration may warrant separate heating, ventilation, and air conditioning (HVAC) systems, but that other l options may be available to the designer. In its SRM of June 26, 1990, the i Comeission approved the staff's position, as supplemented by the staff's response of April 27, 1990. 4
i l i In a letter dated May 5,1992, EPRI stated, "The ALWR program requirements have been specified to provide for three-hour fire barriers between redundant l safety systems and to prevent the migration of smoke and hot gases between 1 compartments containing redundant safety systems by requiring design of l separate HVAC systems to serve redundant trains of safety equipment. i Exceptions to these requirements are the main control room and reactor containment In its letter dated August 17, 1992, ACRS stated, "Except for the concern with shared HVAC, we support the staff recommendation that the passive plants should be reviewed against the enhanced fire protection criteria approved in ! the Commission's SRM." The ACRS concern with shared HVAC relates to the need for adequate isolation of such systems during certain disruptive events (such as fires, floods, or pipe breaks). ACRS stated: If the isolation is not adequate, the HVAC arrangemant may become a pathway whereby effluents from the event are conducted to locations where required safe-shutdown equipment is located. This is not a concern if either (1) the HVAC isolation provisions are able to withstand the ; event consequences (e.g., pipe whip, jet impingement, static and dynamic pressure, and elevated temperature) during and after closure with consideration of single active component failures and acceptable leakage, or (2) the safe shutdown equipment is qualified for the environmental exposure resulting from a release of the adverse environment at any credible location along the HVAC pathway, such as duct openings or blowout locations. The staff maintains that the proposed requirement to consider the effects of smoke, heat, and fire suppressant migration may warrant separate HVAC systems, but that other options may be available to the designer. The EPRI Requirements Document and the evolutionary ALWR designers have indicated that their fire protection designs are consistent with the staff's proposed enhancements. The staff is in the process of evaluating their submittals to ensure acceptable implementation of the Commission's guidance on this issue, and does not expect any related policy matters to result from its review. EPRI has specified requirements for the passive designs similar to those for the evolutionary designs. The passive ALWR vendors have indicated that their designs will comply with the requirements document for passive designs. The staff has not identified any unique features of the passive designs that would preclude the staff's conclusion that these designs should also be evaluated against the enhanced fire protection criteria. Therefore. the staff rec T.ds that the C - ission aoorove the cosition that the oassive plants should also be reviewed acainst the enhanced fire protection criteria anoroved in the Commission's SRM of June 26. 1990.
4
- F. Intersystem Loss-of-Coolant Accident i
In SECY-90-016, the staff recommended that evolutionary ALWR designers should ! reduce the possibility of a loss-of-coolant accident (LOCA) outside j containment by designing (to the extent practicable) all systems and i subsystems connected to the RCS to withstand the full RCS pressure. The staff ) further recommended that systems that have not been designed to full RCS j pressure should include: the capability for leak testing of the pressure isolation valves; i
- valve position indication that is available in the control room when j isolation valve operators are deenergized; and, l
- high-pressure alarms to warn control room operators when rising reactor
- coolant (RC) pressure approaches the design pressure of attached low-
[ pressure systems and both isolation valves are not closed. l In its SRM of June 26, 1990, the Commission approved the staff's position on
- intersystem LOCA, provided that all elements of the low-pressure system are i considered (including instrument lines, pump seals, heat exchanger tubes, and i valve bonnets).
i i The EPRI requirements document and the evolutionary ALWR designers have j indicated that their submittals are consistent with the approved resolution to i this issue. The staff is in the process of evaluating their submittals to j ensure acceptable implementation of the Commission's guidance on this issue, but does not expect any related policy matters to result from its review. EPRI has specified requirements for the passive designs similar to those for i the evolutionary designs. The passive ALWR vendors have indicated that their 1 designs will comply with the applicable requirements document. 4 ! In its letter of May 5,1992, EPRI stated that in order to ensure that the l guidance of SECY-90-016 is applied to all systems and subsystems connecting to j the RCS, the Utility Requirements Document (URD) was revised to include the
- following specific requirements
- All systems and subsystems connected to the RCS which extended outside 4 the primary containment boundary must be designed to the extent
- practicable to an ultimate rupture strength (URS) of at least equal to j full RCS pressure.
The designer must determine by evaluation that, for interfacing systems or subsystems which do not meet the full RCS URS requirement, the degree i and quality of isolation or reduced severity of the potential pressure i challenges are sufficient to preclude an intersystem LOCA. 4
;
- Additional testing and control room alam capabilities must be j implemented to help reduce the probability of an intersystem LOCA.
4
4 1 In general, the staff has found that these requirements are consistent with the staff position. However, as stated in the draft safety evaluation report ! (SER) on the passive URD, it will be necessary for the plant designer to i demon::trate that any interfacing system for which the URS is not at least ' equal to full RCS pressure could not practically be designed to meet such a criterion. The degree of isolation or number of barriers (for example, three isolation valves) is not sufficient justification for using low-pressure components that can practically be designed to the full RCS URS criterion. In addition, piping runs should be designed to meet the full RCS URS criterion, as should all associated elements (such as flanges, connectors, packing, valve stem seals, : vent lines). pump The seals, heat exchanger plant designer should make tubes,every valveeffort bonnets, and RCS to reduce drain and the level i of pressure challenge to all systems and subsystems connected to the RCS. l The staff has not identified any unique features of the passive plant designs that would preclude the staff's conclusion that these designs should also be evaluated against the staff's previous recommendation. Therefore. the staff recomends that the Commission aoorove the oosition that the passive olants should also be reviewed for compliance with the intersystem LOCA criteria anoroved in the Commission's SRM of June 26. 1090. ' i G. Hydrogen Control i l Containments are required to be designed for control of hydrogen generation ! following an accident and 10 CFR 52.47(a)(ii) requires all applicants for design certification to demonstrate compliance with any technically relevant l portions of the Three Mile Island requirements. For example,10 CFR ! 50.34(f)(2)(ix), Additiona? T#I-related requirements, requires a hydrogen control system that can safely accommodate hydrogen generated by the equivalent of a 100-percent fuel-clad metal water reaction. The system must also ensure that uniformly distributed hydrogen concentrations in the containment do not exceed 10 percent (by volume), or that the post-accident atmosphere will not support hydrogen combustion. Because of the uncertainties in the phenomenological knowledge of hydrogen generation and combustion, the staff recommended in SECY-90-016 that evolu-tionary ALWRs should be designed, as a minimum, to satisfy the following criteria: accommodate hydrogen generation equivalent to a 100-percent metal-water reaction of the fuel cladding; limit containment hydrogen concentration to no greater than 10 percent; and, a provide containment-wide hydrogen control (such as igniters or inerting) for severe accidents. I % . _ _ _ _ _ _ . _ _ __ _ _ _ _ _ _ _.__ _- . - _ - - - , , .cw
The staff recommended that the Commission approve its position that the requirements of 10 CFR 50.34(f)(2)(ix) should remain unchanged for evolutionary ALWRs. In its SRM of June 26, 1990, the Commission approved the staff's position. In its letter of December 6,1991, EPRI stated that its requirements document for evolutionary designs will be modified to fully comply with the above y positions. The ABWR design meets the requirements of 10 CFR 50.34(f)(2)(ix) by using, in conjunction with other systems, a nitrogen-inerted containment atmosphere. The ABB-CE System 60+ has a hydrogen mitigation system consisting of igniters to accommodate the hydrogen production from a 100-percent fuel-clad metal-water reaction and maintain the average containment hydrogen cencentration below 10 percent. The staff is in the process of evaluating these submittals to ensure acceptable implementation of the Commission's guidance on this issue but does not expect any related policy matters to result from its review. In a letter dated May 5, 1992, EPRI indicated that the Requirements Document for evolutionary plants had been revised to require use of a containment-wide hydrogen control system that meets the requirements of 10 CFR 50.34(f). In addition, EPRI indicated that changes were being developed to require a hydrogen control system suitable for passive plants, which would also meet the regulations. The ALWR program intends to specify hydrogen control requirements consistent with NRC staff guidance. The passive ALWR vendors have indicated that their designs will comply with the applicable requirements riocument. In its letter of August 17, 1992, ACRS indicated that it supported the staff's recommendation that hydrogen mitigation systems accommodate a 100-percent fuel-clad metal-water reaction. ACRS recommended that the staff perform an analysis similar to that conducted in support of the resolution of Generic Issue 106, "Pipir.g and Use of Highly Combustible Gases in Vital Areas" on the impact of hydrogen combustion and possible detonation including stratifi-cation, before establishing an average hydrogen concentration limit. The staff has performed numerous studies and conducted several experimental programs to better understand the behavior of hydrogen combustion and potential concentration gradients within the primary containment. While it is ; clear that additional analyses would add to the overall data base, the staff be11 eves that a sufficient base exists today to go forwarti with licensing criteria. With respect to the insights gained from Generic Issue 106, the l staff has extensively examined the potential for and consequences of i detonations. In the last decade, the results of these efforts have been ' published in NUREG/CR-4905, 496), 5275, and 5525. The current NRC research program is also examining the generic issue of local explosions and the initiation of detonations by jet flames, i
1 4 The staff recommends that the Commission aoorove the e sition that cassive plants should be desianed. as a minimum. to the same recuirements ano11ed to evolutionary desians. as anoroved by the Commission's SRM of June 26. 1990. Specifically. cassive plant desians must:
- accommodate hydroaen aeneration eauivalent to a 100-oercent metal-water reaction of the fuel claddina:
- limit containment hydrocen concentration to no areater than 10 percent:
REL.
- orovide containment-wide hydrocen control (such as ianiters or inertina) for severe accidents.
.H . Core Debris Coolability In the unlikely event of a severe accident in which the core melts through the reactor vessel, it is possible that containment integrity could be breached if the molten core is not sufficiently cooled. In addition, interactions between the core debris and concrete can generate large quantities of additional hydrogen and other non-condensible gases, whkh could contribute to eventual overpressure failure of the containment. Therefore, the staff concluded that plant designs should include features to enhance core debris coolability.
In SECY-90-016, the staff recommended that the Commission approve the general criteria that evolutionary ALWR designs:
- provide sufficient reactor cavity floor space to enhance debris spreading; and, -
provide for quenching debris in the reactor cavity. In its SRM of June 26, 1990, the Commission approved the staff's position. In addition, the staff indicated in SECY-90-016 that it was evaluating the level of protection afforded by covering the containment liner and other structural members with concrete. Debris coolability is an area in which there is active ongoing experimental research including relatively large scale testing jointly sponsored by EPRI and NRC. The results of tests were expected to demonstrate the early quenchability of core debris within the reactor cavity. However, these tests were indeterminate in proving quenchability and l indicated the need to consider the potential for continued core-concrete interaction. Concrete can be used as a sacrificial barrier, for both the , liner and structural components, to accommodate potential longer periods of core-concrete interaction. The staff now concludes that it may be necessary to protect these structural components with concrete. The EPRI requirements document and the evolutionary ALWR designs provide a number of design features that are intended to mitigate the effects of a molten core. Among other features, the evolutionary designs proposed floor ___________ _ )
l sizing criteria of 0.02 m*/MWt, and provisions to flood the lower drywell or i reactor cavity. Tpe staff neither supports nor disputes the EPRI floor sizing criteria of 0.02 m /Mwt. Instead, the staff concludes that it is appropriate to review the specific vendor designs to determine how they address the general criteria discussed above (including protecting structural components
- with concrete) to provide an increased level of protection relative to core i debris coolab111ty. The staff concludes that the " core-on-the-floor" accident i will not be considered as a new design-basis accident. However, the staff
- expects'the vendors to consider the effects of core-concrete interactions on I i the production of non-condensible gases, the release of additional fission i j products froci the core-concrete interaction, and additional heat and hydrogen j generation in the new designs.
t The criteria discussed above are intended to ensure that the ALWR vendors l provide measures (to the extent practical) to mitigate severe accidents, while i avoiding turning severe accidents into traditional design-basis accidents. j Because the staff neither supports nor disputes particular floor sizing j criteria, vendors should ensure that the containment can withstand the j environmental conditions (pressure and temperature) and structural challenge j caused by core-concrete interactions. For the range of severe accidents of I concern, vendors should realistically estimate the amount of core-concrete interaction that will occur, and ensure that the containment will accommodate , the resultant conditions for approximately 24 hours without loss of ! containment integrity. Where insufficient data exist to develop realistic
- estimates, vendors may propose alternatives (such as additional tests or the
- use of other methodologies) for determining the degree of core-concrete
! interaction. The ALWR vendors should also perform parametric studies to i determine how sensitive the containment response is to variations in the l amount of core-debris available to interact with the concrete. 1 j The staff concludes that incorporation of mitigative measures (to the extent practical) and assurance of containment integrity for approximately 24 hours ! will provide defense-in-depth as well as an appropriate degree of robustness in the containment design. ] ! In its letter of May 5,1992, EPRI indicated that the requirements documents
! specify requirements to address debris coolability including cavity / lower drywell area to permit spreading, cavity / lower drywell flooding to quench debris, and protection of the containment boundary from debris attack. The requirements documents also specify that containment loads from dominant core damage sequences be evaluated. The passive ALWR vendors have indicated that their designs will comply with the applicable EPRI requirements document.
The staff recommends that the Commission accrove the position that both the evolutionary and cassive LWR desions meet the followino criteria:
- orovide reactor cavity floor space to enhance debris screadina:
e provide a means to flood the reactor cavity to assist in the coolina Drocess: 4 1 I l d
i j - . l l
- _ 14 -
i j . orotect the containment liner and other structural =a=hers with concrete. if necessary: and. i ensure that the best estimate environmental conditions fore'ssure and tannerature) resultina from core-concrete interactions do not exceed
- Service Level C for. steel containments or Factored Load Cateaory for i concrete containments. for anoroximately 24 hours. Ensure thatJ.he containment canability has marain to accommodate uncEtaintie
- ; in the
- j. environmental conditions from core-concrete interacticns. _
I. High-Pressure Core Melt Ejection In SECY-90-016, the staff recommended that the Commission approve the position that evolutionary ALWR designs should include a depressurization system and cavity design features to contain ejected core debris in order to reduce the potential for containment failure as a result of direct containment heating (DCH). The staff is concerned that this event might result from the ejection of molten core debris under high-pressure from the reactor vessel. Such an ejection might result in wide dispersal of core debris, rapid oxidation, and extremely rapid addition nf energy to the containment atmosphere. In its SRM of June 26, 1990, the Commission approved the staff's position. The Commission also directed that the cavity design, as a mitigating feature, should not unduly interfere with operations, including refueling, maintenance, or surveillance activities. Examples of cavity design features that will decrease the amount of ejected core debris that reaches the upper containment include (1) ledges or walls that would deflect core debris and (2) an indirect path from the lower reactor cavity to the upper containment. The staff will review the LWR designs relative to the above criteria. In its letter of May 5,1992, EPRI indicated that the requirements document specifies an RCS depressurization system and cavity retention capability for both evolutionary and passive plants. EPRI further indicated that since the passive plant emergency core cooling system (ECCS) relies on RCS depressurization, redundancy and diversity have been specified for the depressurization system to ensure very high reliability. In its letter of August 17, 1992, ACRS indicated that because direct containment heating is an extremely improbable event, two modes of coping with the possibility are not needed. ACRS stated that because of the possible safety benefits for other events, reliable depressurization is the preferred approach. The staff agrees with'the ACRS assertion that a reliable depressurization system is needed. However, the staff proposes to provide a design concept with a degree of consequence mitigation along with a certain amount of accident prevention. The depressurization system retains a degree of uncertainty. Such questions as the rate of depressurization, the timing for i
operator initiation of manual depressurization, and the cut-off pressure may never be totally resolved. As a result, the staff believes that a design can be developed to decrease the direct flight path to the upper containment at little or no added expense. The plant designers have proviced features to address this issue for evolutionary ALWR Jesigns. The staff is in the process of evaluating their l submittals to ensure acceptable implementation of the Commission's guidance on ; this issue. The staff's preliminary review of the passive ALWRs has also l identified the importance of RCS depressurization to the safe shutdown of the l plant during transients or accidents. RCS depressurization is crucial to the operation of the passive safety features that limit the likelihood of core damage, as well as to reducing the potential for containment failure by direct , containment heating from the ejection of core debris at high pressure. I Therefore, the staff has determined that the passive ALWR designs should 1 include a highly reliable depressurization systen. ) The staff recommends that the Commission aoorove the aeneral criteria that the evolutionary and passive LWR desian_i_
- orovide a reliable deoressurization syst9m: and. ,
- orovide cavity desion features to decrease the amount of eiected core debris that reaches the unoer containment, i J. Containment Performance !
As discussed in SECY-90-016, the staff recommended that the Commission approve the position to evaluate evolutionary ALWRs using a conditional containment failure probability (CCFP) of 0.1, or a deterministic containment performance l goal that offers comparable protection. The staff concluded that the ' following general criterion would be an appropriate substitute for a CCFP in evaluating evolutionary ALWR ccatainment performance during a severe-accident challenge: The contr.inment should maintain its role as a reliable, leak-tight barrier by ensuring that containment stresses do not exceed ASME service level C limits for a minimum period of 24 hours following the onset of core damage, and that following this 24-hour period the containment should continue to provide a barrier against the uncontrolled release of l fission products. I l The staff proposed this containment perfomance goal to ensure that the - l containment will perform its function in the face of most credible severe accident challenges. In its SRM of June 26, 1990, the Commission approved the use of a 0.1 CCFP as a basis for establishing regulatory guidance for the evolutionary ALWRs. The l Commission directed, however, that this objective should not be imposed as a i requirement, and that the use of the CCFP should not discourage accident i 1
1 4 4 i j ! prevention. j The Commission also directed the staff to review and submit to the Commission suitable alternative, deterministically established containment i perfomance objectives providing comparable mitigation capability that may be considered by the applicants. i In this paper and in SECY-90-016, the staff has identified the major i challenges to the containment (such as hydrogen burns or core debris ! interactions with water and containment structures) and the need to provide i the means for mitigation of these challenges. Nonetheless, the containment performance goal acts as defense-in-depth to ensure that the design (including i its mitigation features) would be adequate if called upon to mitigate a severe j accident. Although not explicitly identified in SECV-90-016, the staff will also ! evaluate the impact of interaction between molten fuel and coolant, and the resulting steam and hydrogen generation (and any dynamic forces due to ex- ) vessel fuel-coolant interactions outside the vessel) on the integrity of the i containment, consistent with the containment performance goal. The evaluation j of containment bypass sequences will be addressed on a vendor-specific basis j during the staff's review of ALWR designs. s , l The intent of both the CCFP and the alternative deterministic performance i criteria discussed above is to provide a final check as well as defense-in- ! depth. The philosophy behind the use of the proposed deterministic goal is that adequate time must be provided for fission product decay before allowing j
~
a release from the containment to the environment. l 'n its letter of August 17, 1992, ACRS stated that the staff has not developed j an adequate technical position relating to requirements for containment i performance in passive designs. ACRS also contends that the CCFP approach j should be used in developing reguiatory requirements and not merely passed on j to the applicants. With regard to deterministic evaluation of cantainment performance, ACRS implied that the staff should identify the containment challenges, or how they are to be quantified, on a generic basis. ACRS l endorsed the criteria recommended in its letter of May 17, 1992, " Proposed = Criteria to Accommodate Severe Accidents in Containment Design." I During the evolutionary ALWR reviews, the staff conducted a thorough review to ensure that a probabilistic CCFP would not be used in a way that could detract from a balanced approach of severe accident prevention and consequence mitigation. The ABWR review is nearly complete and, at this time, the staff believes GE has met the CCFP goal of 0.1 for internal events. The AB8-Cf System 80+ design is also expected to meet the CCFP goal of 0.1. Through the review of the evolutionary ALWR designs, the staff recognized the limitations of a CCFP approach. Specifically, as accident prevention is increased by decreasing core damage frequency, the ability of the containment to withstand events of even lower probability becomes less dear. The limitations of a CCFP approach are also evident in the uncertainties prevalent ' within a probabilistic risk assessment (PRA). Because of these limitations,
j i i 4 i vendors have provided some deterministic analysis to complement the CCFP ] approach by addressing uncertainties in the severe accident phenomena and
- calculation of the CCFP. In particular, uncertainties surrounding the issue of debris coolability have led the staff to conclude that deterministic i best-estimate analyses should complement the CCFP approach. In spite of its limitations, the staff believes that the CCFP approach ensures that i evolutionary ALWRs maintain a balance between accident prevention and 1 consequence mitigation.
In its letter of May 5, 1992, EPRI indicated that the requirements document specifies a set of deterministic contair, ment performance requirements
- supported by PRA. These requirements address a set of severe accident 4 containment challenges and specify that ASME Service Level C limits be met i for risk-significant sequences (expected to be low-pressure core melt into an
- intact containment). EPRI indicated that the proposed staff criteria are 4
generally consistent with the ALWR requirements, except in the challenges to
- be considered in demonstrating the 24-hour. Service Level C goal. EPRI has provided a containment performance study for both evolutionary and passive J
plant designs that identifies 23 postulated containment challenges and failure modes along with specification of design features for their mitigation and a requirements for deterministic analysis. The staff has reviewed the listing of identified challenges and believes that it represents a complete set. i In SECY-92-292, " Advance Notice of Proposed Rulemaking on Severe Accident Plant Performance Criteria for Future ALWRs," dated August 24, 1992, the staff
- proposed that the Commission issue an Advance Notice of Proposed Rulemaking (ANPR) on Severe Accident Plant Performance Criteria for Future LWRs.
Intended to be used for passive LWRs, the ANPR contains three distinct options that could be followed to establish severe accident containment performance l requirements. The three options include a hardware-oriented rule, a phenomena-oriented rule, and a general design criteria (GDC)-oriented rule that was outlined in detail in an ACRS letter dated May 17, 1991. In an SRM dated September 17, 1992, the Commission directed the staff to place the ANPR in the Federal Reaister for a 90-day comment period. The staff will
- review the containment performance goal and severe accident guidance following i receipt of all public comment and after meeting with Westinghouse and GE representatives to discuss their passive LWR designs. The staff will also discuss these issues with ACRS.
l The staff is currently evaluating the containment performance of the individual ALWR designs to ensure that all potential sequences that may be identified during the staff's review are adequately addressed. The staff will evaluate the criteria used by the vendor to determine the challenges to the containment. l In SECY-90-016, the stzff indicated that a general goal of limiting the i conditiont.1 containeeM failure probsollity to less than 1 in 10 when weighted j over credible core-damage sequences would constitute appropriate attention to the defense-in-depth philosophy. Alternatively, a deterministic containment
- - - . - - .m.-- .,.,y , _ - - _ , _ ,,
l i j j l $ performance goal that provides comparable protection would be appropriate. ' 1 8ecause both containment performance goals are acceptable and given the i limitations inherent within the CCFP approach, the staff believes the i deterministic containment performance goal should be pursued for the passive i ! ALWR designs. l i
- Although the staff is still evaluating this issue and expects further insight !
j from public comments on the ANPR, the staff concludes that it is appropriate l
- to proceed with interim severe accident containment performance criteria.
] These interim criteria are intended to reflect the lessons learned during the l j' review of the evolutionary designs. The staff recommends that the Commission anorove the oosition to use the followino deterministic containment
- cerformance coal in the evaluation of the passive ALWRs
The containment should maintain its role as a reliable. leak-ticht J barrier (for examole. by ensurino that containments stresses do not exceed ASME Service Level C limits for metal containments. or i Factored Load Cateaory for concrete containments) aooroximately 24 hours fol lowino the onset of core damaae under the more likely i severe accident challences and. followino this ceriod. the j containment should continue to orovide a barrier naainst the ] uncontrolled release of fission products. l t i The staff will inform the Commission if it identifies additional policy issues 4 as a result of (1) reviewing ALWR designs, or (2) evaluating comments on the i ANPR concerning " Severe Accident Plant Performance Criteria for Future LWRs." ! K. Dedicated Containment Vent Penetration ! In SECY-90-016, the staff recommended that the Commission approve the use of , j an overpressure protection system that uses a dedicated containment vent for I the ABWR. This system should be designed to avoid gross containment failure ! resulting from postulated slow rising overpressure scenarios that could result
- from postulated multiple safety system failures. '
s ' In its SRM of June 26,' 1990, the Commission approved the staff's recommended ! use of the containment overpressure protection system for the ABWR, subject to t ! a comprehensive regulatory review to consider the "downside" risks with the l j mitigation benefits of the system. In addition, the Commission directed the i staff to ensure that the design provides full capability to maintain control j over the venting process. In its letter of May 5, 1992. EPRI' indicated that containment venting was an i optimization issue in the EPRI requirements documents. According to the ALWR 1 i Utility Steering Committee representatives, specific containment overpressure j protection for evolutionary ALWR designs, either through the size and strength of containment or through installation of an overpressure protection relief, j is considered an acceptable approach. 2 i
j . - i j , a l l EPRI indicated that passive plant design features that address the containment I { overpressure challenge include highly reliable, redundant, and diverse passive i safety-grade decay heat removal, automatic depressurization, and containment j cooling. EPRI recommended that the NRC not require a containment vent in < j evolutionary PWR containments and passive ALWRs. l In its letter of August 17, 1992, ACRS indicated that the Commission should i make a generic judgement about the acceptability of containment vents for light water reactors. ACRS contends that this should be part of establishing l
! general criteria for containment designs, as proposed in their letter of i
! May 17, 1992. 1 i Tb staff considers the containment vent as one of many plant systems that can ! be used to mitigate the consequences of an accident. If acceptable analyses
' indicate that a vent would not be needed to meet the severe accident criteria, i such as the containment performance goal, the staff would not propose to l implement a vent requirement.
k' Because of the current stage of design development and review, the staff has 1 insufficient information at this time to determine whether a containment vent ' is necessary for passive plant designs. The containment performance criteria
- proposed in Section 1.J of this enclosure will serve as the basis for the
- staff's review of containment integrity and the need for a containment vent.
Therefore. the staff recommends that the Commission anorove the cosition that l the need for a containment vent for the cassive clant desions should be j evaluated on a desion-soecific basis. L. Equipment Survivability 4 , In SECY-90-016, the staff recommended that the Commission approve the position
, that features provided only for severe-accident protection need not be subject to the environmental qualification requirements of 10 CFR 50.49; quality assurance requirements of 10 CFR Part 50, Appendix B; or redundancy / diversity requirements 10 CFR Part 50, Appendix A. The reason for this judgement is
! that the staff does not believe that severe core damage accidents should be
; treated in the same manner traditionally used for design-basis accidents
] (DBAs) because of significant differences in their likelihood of occurrence. However, SECY-90-016 further stated that mitigation features must be designed to provide reasonable assurance that they will operate in the severe-accident , j environment for which they are intended and over the time span for which they l are needed. In instances where safety-related equipment provided for DBAs is relied upon to cope with severe-accident situations, there should also be a j high confidence that this equipment will survive severe-accident conditions for the period that it is needed to perform its intended function. i
- During the review of the credible severe-accident scenarios for ALWR designs, I the staff will evaluate the ALWR vendors identification of the equipment I 4 needed to perform mitigative functions and the conditions under which the '
mitigative systems must operate. Equipment survivability expectations under f severe-accident conditions should consider the circumstances of applicable 2 4 n - ~ m --- m --
! l 1 i l f l } ! l initiating events (such as station blackout or earthquakes) and the ! environment (including pressure, temperature, and radiation) in which the ! equipment is relied upon to function. The required system performance l criteria will be based on the results of these design-specific reviews.
- In its SRM of June 26, 1990, the Commission approved the staff's position. In i its letter of May 6,1991, the staff clarified its position tiat these j criteria would be applied to features provided only for severe accident i mitigation.
The EPRI requirements document and the evolutionary ALWR designers have indicated that their submittals are consistent with these criteria. The staff is in the process of evaluating their submittals to ensure acceptable imple- 1 mentation of the Commission's guidance on this issue, but does not expect any related policy matters to result from its review. The passive ALWR vendors have indicated that their designs will comply with the applicable EPRI requirements document. In its letter of August 17, 1992, ACRS agreed with the staff position discussed above. Although the staff is still evaluating this issue for the passive plant designs, it has not identified any unique features of the designs that would preclude the staff's conclusion that these designs sinould also be evaluated against the criteria established for evolutionary plants. Therefore. the staff recommends that the Commission aoorove the oosition that cassive olant desian features orovided only for severe-accident mitiaation need not be subiect to the environmental cualification reouirements of 10 CFR Section 50.49: cuality assurance reouirements of 10 CFR Part 50. Annendix B: and redundancv/ diversity of reouirements 10 CFR Part 50. Anoendix A. As discussed in SECY-90-016, the staff concludes that guidance such as that found in Appendices A and 8 of RG 1.155, " Station Blackout," is appropriate for equipment used to mitigate the consequences of severe accidents. M. Elimination of Operating-Basis Earthquake In SECY-90-016, the staff' discussed its proposal to decouple the operating-basis earthquake (OBE) from the safe-shutdown earthquake (SSE). The reguia-tions in Appendix A to 10 CFR Part 100 establish the OBE at one-half the intensity of the SSE. The staff stated that the OBE should not control the j design of safety systems and was evaluating possible changes to the ' regulations that would reduce the magnitude of the 08E relative to the SSE. The staff recommended that the Commission approve the review approach to ! consider requests to decouple the OBE from the SSE on a design-specific basis i for evolutionary designs. In its SRM of June 26, 1990, the Commission approved the staff's recommendation. j In a plant optimization subject, EPRI requested that the staff evaluate the total elimination of the OBE from the design of systems, structures, and components (SSC) in nuclear power plants. In its letter of April 26, 1990, the ACRS also recommended this approach. In evaluating the decoupling of the OBE from the SSE, the NRC staff is also evaluating the possibility of i
i 1 , I redefining the OBE in order to satisfy its function without an explicit response analysis. This change would diminish the role of the OBE in design by establishing a level which, if exceeded, would require that the plant be
- shut down for inspection activities.
! EPRI's position on seismic. design is that it is unnecessary to perform two i complete sets of seismic analyses -- one for the OBE and one for the SSE. The ! NRC staff agrees, in principle, with this position, but finds that extant i design practices for piping and structures do not result in designs that are ; i significantly controlled by the OBE. 'As stated in SECY-90-016, certain l interim measures (such as allowing higher damping values for piping analyses) i have already been implemented to mitigate the situation of having the OBE i ! significantly control the design. i )
- The elimination of the OBE response analysis would require performance of all current OBE design-related checks for a fraction of the SSE. The staff is l working with the industry to develop various design alternatives for the ABWR j i to supplement the codes and standards when design-related checks are based on 3 the OBE. For example, in piping design, the ASME Boiler and Pressure Vessel Code currently establishes rules for evaluating earthquake cycles on fatigue and relative seismic anchor motion effects that are based on the OBE. The
- staff and industry are supplementing the Code with design rules that would f account for fatigue and seismic anchor motion effects based on the SSE. In addition, the NRC guidelines for postulating the number and location of pipe ruptures are also derived from the OBE. When the OBE is eliminated from design, these loading calculations may need to be performed using the SSE after establishing appropriate new allowable limits.
The staff's proposed amendment of Appendix A to 10 CFR Part 100 would allow the option to eliminate the OBE from design certification when the OBE is i established at less than or equal to one-third the SSE. In this manner, the 1 03E serves as an " inspection level earthquake" below which the effect on the health and safety of the public would be insignificant and above which the licensee would be required to shut down the plant and inspect for damage. The staff assessed the safety margins of several aspects of nuclear plant design when the OBE is eliminated from consideration. The industry and staff recognize that earlier seismic criteria caused the OBE to control certain aspects of the plant design (such as the piping systems). The industry and ' the staff view the " controlling" nature of the OBE design as an additional margin above the safety margins established by the design bases. Therefore, eliminating the OBE would not result in a significant decrease in the overall plant safety margin. The staff is currently performing a detailed evaluation of the extent to which the OBE controls the design and the effect on the design of SSC when the OBE is eliminated from design consideration. The overall design of reactor site . structures is generally conservative, and the structural responses for all combinations of loads (including those from
, earthquakes) are kept at or below the material yield stresses to preclude plasti: deformation.
i The staff has examined the structural load combinations and the corresponding acceptance criteria. On the basis of analyses, tests, and engineering judgment, the staff has determined that the structural design produced using SSE load combinations envelop the load combinations produced using the 08E. These conclusions are consistent with the staff's licensing experience accumulated during its review of many seismic calculations for individual plants and test data from NRC-sponsored research. For analyses of safety-related structures, the effects of the relative displacements between adjacent structures need to be considered for earth-quakes. With the elimination of the OBE, these effects should be considered for the SSE and would also be needed for input into the piping design as discussed later in this issue. In RGs 1.27 and 1.143, the staff recommends that the designs of the ultimate heat sink and radioactive waste structures, respectively, be evaluated only for OBE loading using the appropriate load combinations and limits for the OBE. When the OBE is eliminated from the A8WR, the staff found that GE proposed to design these buildings and structures to the SSE loadings. The staff concluded that using SSE loadings and the appropriate load combinations and limits provides a bounding design comparable to that provided in the regulatory guides. The staff will also review alternative methods that might be proposed by other vendors on a case-by-case basis. 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). After reviewing these aspects, the staff 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 dominates the load combination. Therefore, the staff concludes that eliminating the OBE piping stress load combinations for primary stresses in piping systems will not significantly reduce the 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 08E 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 08E loadings. Accordingly, to account for earthquake cycles in the fatigue analysis of piping systems, the staff proposes to develop guidelines for selecting a number of SSE cycles at a fraction of the peak amplitude of the SSE. These guidelines will provide a level of fatigue design for the piping equivalent to that currently provided in the standard review plan (SRP)(NUREG-0800). Currently, the staff's guidelines in SRP Section 3.9.2 recommend an equivalent
j t of 50 OBE peak cycles for fatigue evaluation. The staff will develop new guidelines after conducting regulatory research and will incorporate the , guidelines into an SRP revision or into a regulatory guide, as necessary. j To account for earthouake cycles in the faticue analyses of oioina systems
- nerformed until the new outdance is issued. the staff crocoses usina two SSE j events with 10 maximum stress cycles ner event (20 full cycles of the maxim =
- SSE stress ranae). This is equivalent to the cyclic load basis of one SSE and i five OBE events, as currently recommended in SRP Section 3.9.2, when i accounting for differences in the structural damping between the OBE and SSE and for a 60-year (instead of a 40-year) plant life. Alternatively. the staff nronoses that number of fractional vibratory cycles eauivalent to that of 20 full SSE vibratory cycles may be used (but with an amolitude not less than one-third of the maximum SSE amolitude) when derived in accordance with !
Anoendix 0 of IEEE Standard 344-1987. il The ASME Boiler and Pressure Vessel Code (ASME Code), Section III, Paragraph
- NC/ND-3655 specifies that seismic anchor displacement effects need not be j considered for Service Level D. However, the ASME Code requires that seismic anchor motion stresses be considered for Service Level B, for which the OBE has_ traditionally been the designated seismic loading. If the OBE were eliminated from the piping design, the ASME Code, Section III evaluation would
- have no requirement for considering the effects of seismic anchor motion. Ihg
- staff oronoses that the effects of anchor disolacements in the einina caused by an SSE be considered with the Service Level D limit. The staff's
- recommendation will correct this anomaly and will require an evaluation of 4
seismic anchor motion effects for the SSE together with the effects of normal
- conditions as required by 10 CFR Part 50, Appendix A, GDC 2. Their effects would be evaluated to a Service Level D limit for which the SSE has traditionally been the designated seismic loading.
The staff proposes that existing staff guidelines ensuring the functionality . of safety-related components and supports under SSE loading conditions be l maintained. For example, when safety-related equipment is qualified by analysis only, the stress limit should remain in the linear-elastic range for SSE loading. Similarly, the function of the supported system must be taken . into account. As specified in RG 1.124, " Service Limits and Loading Combina-tions for Class 1 Linear-Type Components Supports," the ASME Service Level 8 limits of Subsection NF (or other justifiable limits approved by the staff) should be used to ensure that systems which normally prevent or mitigate ' consequences of SSE-related events will operate adequately regardless of plant condition. Pipe rupture is a rare event that can be caused by errors in design, con-struction, or operation; unanticipated loads; or unanticipated corrosive environments. The staff notes that piping failures generally occur at high stress and fatigue locations, such as at the ends of a piping system where it connects to component nozzles.
i - 1 1 j 4 Recent dynamic pipe tests conducted by EPRI and the NRC demonstrated that
- butt-welded piping can withstand seismic inertial loadings higher than an SSE j without rupturing. Thus, the staff concludes that the likelihood of a double-
! ended pipe rupture caused by an OBE-level earthquake in a piping system i designed to an SSE is remote. Operating experience has shown that pipe failures. (splits, through-wall cracks, and double-ended pipe ruptures) are more likely to occur under conditions caused by normal operation. These 1 conditions include erosion, corrosion, thermal constraint, fatigue, and j operational transients. The staff recommends that the Commission aoorove the anoroach to eliminate the
- OBE from the desian of systems structures. and comoonents. When the OBE is I
eliminated from the desion. no reolacement earthouake loadino should be used to establish the oostulated nine ruoture and leakaae crack locations. The ! staff recommends that the criteria for postulating pipe ruptures and leakage cracks in high- and moderate-energy piping systems be based on factors j attributed to nonnal and operational transients alone. However, for J establishing pipe breaks and leakage cracks due to fatigue effects, j calculation of the cumulative usage factor should continue to include seismic ! cyclic effects. a i Further reduction in the nun 6er of postulated pipe rupture locations can be considered when compensatory measures are established to minimize the > potential for pipe ruptures during normal operating and transient conditions 1' i (such as control of erosion / corrosion or use of upgraded piping materials). ? The guidelines for environmental qualification and compartment pressurization
- are currently based on the mechanistic break locations. 7herefore. the staff l oronoses that the mechanistic oine break and hich-enerav ' eakaoe crack i locations detenmined by the cioina hiah stress (without the OBE) and fatique i
locations may be used for eautoment environmental cualification and { connartment oressuriution ourooses. ! Eliminating the OBE from explicit design consideration affects several aspects of the seismic qualification of safety-related mechanical and electrical i ' 4 equipment. When the equipment qualification is performed by analysis, the i acceptance criteria are derived from the ASME Code. The effect of eliminating l the OBE from equipment qualification by analysis should be negligible. It is i 4 well known that mechanical equipment (such as pumps and valves) is, in . general, seismically rugged when adequately anchored. It is also known that operability limits for mechanical equipment are generally established through maximum permissible moments and forces or tolerance limits based on available clearances that are controlled by the SSE (rather than the 08E). Therefore, for mechanical equipment, elimination of OBE from qualification by analysis should not reduce any margin. Also, some electrical equipment may be 3 qualified by an analysis requiring demonstration that five 08E events followed j by one SSE event do not cause a failure of the equipment to perform its safety j function. With the elimination of OBE, analysis checks for fatigue effects i can be performed at a fraction of the SSE (such as 50 cycles at one-half of } the SSE peak amplitude, or 150 cycles at one-third of the SSE peak amplitude). 3 } I ?
t l } i i l 1 When equipment qualification for seismic loadings is performed by analysis, ! testing, or a combination of both, the staff recommends the use of IEEE Standard 344-1987, as endorsed in RG 1.100, Revision 2. For such analysis,
- . the selection of the service limit level for different loading combinations will ensure the functionality of the equipment during and following a SSE.
! For testing, IEEE Standard 344-1987 details requirements for performing seismic qualification using five OBE events followed by an SSE event. Where
- complex mathematical models are based solely on calculated structural j parameters, verification testing should be performed.
~; With the elimination of the OBE. two alternatives exist that will essentially ' maintain the reouirements provided in IEEE Standard 344-1987 to cualify i eouioment with the eauivalent of five OBE events followed by one SSE event j (with 10 maximum stress cycles oer event). Of these alternatives. the staff concludes that eauipment should be cualified with five one-half SSE events j followed by one full SSE event. Alternatively. a number of fractional peak ! cycles eouivalent to the maximum peak cycles for five one-half SSE events may i be used in accordance with Anoendix 0 of IEEE Standard 344-1987 when followed ] by one full SSE. The staff will conduct research to verify the number of SSE ' cycles and the fraction of their peak amplitude for which the equipment is to be tested. The staff will also review the results of research conducted by l the industry standards group, as appropriate. Lastly, the design of ALWRs using a single-earthquake (that is, SSE) design is predicated on the adequacy of pre-earthquake planning and post-earthquake : , damage inspections that are to be implemented by the combined operating l license (COL) applicant. The staff proposes that the COL applicant submit to : the NRC staff, as a part of its application, the procedures it intends to use l for pre-earthquake planning and post-earthquake inspections. The staff is ) currently developing a regulatory guide for pre-earthquake planning and post- 1 earthquake operator actions. l In its letters dated May 5, 1992, and August 21, 1992, EPRI representatives l commented on the staff's positions concerning eliminating the OBE from design as discussed in the two draft Conmiission papers. EPRI agrees with the staff'r recommendation to eliminate the OBE from design. The EPRI requirements documents will require a seismic margins assessment, which demonstrates a margin for an earthquake substantially larger than the SSE. However, EPRI does not fully agree with the number of earthquake cycles to be used in fatigue evaluation and equipment qualification. The EPRI recommended that the NRC guidelines be similar to those contained in paragraph N-1214 of ASME Code, , Section III, Appendix N, which provides that no more than two OBE events with I a total of 20 full stress cycles (that is, 20 cycles at one-half SSE) be used. Similarly, in its letter dated September 17, 1992, Westinghouse Electric Corporation representatives commented on the staff's position regarding the elimination of the OBE and noted that the staff's proposed position was overly conservative in the number of earthquake cycles to be considered. Westinghouse proposed to adopt the guideline of 20 cycles at one-half of the SSE response, as specified by the EPRI URO for passive plants. I
4 - 4 i The staff's evaluation indicates, however, that the EPRI and Westinghouse j recommendations provide less stringent design requirements than those
; currently prescribed by the staff guidelines of SRP Section 3.9.2.
i Additionally, the current staff guidelines were used for the review of nuclear i plants with a 40-year life or less. 4 i The staff positions presented above have been revised since the draft i Commission papers to account for both a 60-year plant life and the differences i in the structural damping used for the OBE and SSE. (Component damping was 1 not specified because EPRI pointed out that is was the same for the OBE and i SSE in ASME Code Case N-411). The staff also found that the overall contribution of earthquake cycles to fatigue is small. To design for 20 full j SSE cycles (or the equivalent) would not significantly penalize the design and ! would provide a bounding design for the expected number of earthquakes of a j lesser magnitude than the SSE and their aftershocks for a 60-year plant life. In its letter of September 16, 1992, ACRS stated that it believes the staff l took an appropriate approach in its interim position (which has been j incorporated into the final staff position above). J In its final position, the staff supplemented its preliminary positions from j the draft Connission papers and provided a more complete package identifying j the actions necessary for the design of SSC when the OBE is eliminated. As
- discussed above, the staff clarified that guidelines should be maintained to ensure the functionality of components, equipment, and their supports. In i addition, the staff clarified how certain design requirements are to be
- - considered for buildings and structures that are currently designed for the OBE earthquake, but not the SSE. Also, the staff addressed how pre-earthquake
- planning and post-earthquake operator actions are to be considered. The i
staff's proposed guidelines and their bases are discussed in the final ! positions above. } The staff has evaluated the effect on safety of eliminating the OBE from the 4 design load combinations for selected SSC and has developed proposed criteria for an analysis using only the SSE. The staff. therefore. reouests that the j Commission anorove the cronosed oositions discussed above. The staff will ! keep the Commission informed as the review progresses and will note in case- ] specific safety evaluations instances in which the applicant proposes to use l _ criteria different than those described above for an SSE-only analysis. } N. Inservice Testing of Pumps and Valves I i In SECY-90-016, the staff recommended that the Commission approve the position that the following provisions should be applied to all safety-related pumps and valves, and not linited to ASME Code Class 1, 2, and 3 components: i Piping design should incorporate provisions for full flow testing j (maximum design flow) of pumps and check valves. 1 i 4
4 l
- Designs should incorporate provisions to test motor-operated valves j under design-basis differential pressure.
=
! Check valve testing should incorporate the use of advanced, i non-intrusive techniques to address degradation and performance
- characteristics.
L A program should be established to determine the frequency necessary for ! disassembly and inspection of pumps and valves to detect unacceptable j~ degradation that cannot be detected through the use of advanced, non-intrusive techniques. i l The staff concluded that these requirements are necessary to provide an j adequate assurance of operability. l !a its SRM of June 26, 1990, the Commission approved the staff's position as
- supplemented in the staff's response to ACRS comments,. dated April 27, 1990.
- In that response, the staff agreed with the ACRS recommendations to emphasize
- the requirements of Generic Letter (GL) 89-10 with regard to evolutionary i plants, to resolve check valve testing and surveillance issues, and to indi-l cate how these requirements are to be applied to evolutionary plants. The 4
staff also agreed that the requirements should permit consideration of pro-1 posed alternative ways of meeting inservice and surveillance requirements. The Commission further noted that due consideration should be given to the i practicality of designing testing capability, particularly for large pumps and j valves. i j After reviewing the proposed staff requirement in SECY-90-016 which stated i that designs should incorporate provisions to test motor-operated valves under i design-basis differential pressure, the staff has clarified its position. The l i staff recommends that full flow testing be conducted at maximum design flow I f with analysis to extrapolate to design pressure if it is not practicable to conduct the inservice pump testing at design flow and pressure. The staff also recommends that for valves, a qualification test (under design-basis i differential pressure) be conducted prior to installation and inservice valve tests be conducted under the maximum practicable differential pressure and flow when it is not practicable to achieve design-basis differential pressure during an inservice test. In its letter of May 5,1992, EPRI stated that the ALWR program agrees with the above staff positions for the passive and evolutionary plants. In its letter of August 17, 1992, ACRS stated that they support the staff's recommendation that the above design, testing, and inspection peovisions should be imposed on all safety-related pumps and valves for passive ALWRs. ! EPRI and the evolutionary ALWR designers have indicated that their submittals are consistent with these criteria. The staff is in the process of evaluating their submittals to ensure acceptable implementation of the Commission's guidance on this issue. h passive ALWR vendors have indicated that their designs will comply with the applicable EPRI requirements document.
Therefore. the staff reccamonds that the Commission anorove the nosition that these reoutrements shou' c a'so be innosed on nassive ALWRs. The staff concludes that additional inservice testing requirements may be necessary for certain pumps and valves in passive plant designs. This necessity arises because the passive safety systems rely heavily on the proper operation of , this equipment (such as check valves or depressurization valves) to mitigate I the effects of accidents and to shut down the reactor. The staff will discuss l its proposed resolution of this issue in a separate Commission paper ' addressing the regulatory treatment of nonsafety systems in passive plant designs. ~ II. Other Evolutionary and Passive Design Issues A. Industry Codes and Standards In SECY-91-273, " Review of Vendors' Test Program to Support the Design Certification of Passive Light-Water Reactors," dated August 27, 1991, the staff raised the concern that a number of design codes and industry standards dealing with new plant construction have recently been developed or modified, and that the NRC has not yet detemined their acceptability. The staff recommends that the Commissiin anorove the nosition. consistert with nast nractice. that it wil' review bot a evoit tionary an<f nassive n' ant cesian ano11catioms usina the nmmst codes and stancards ttat aave t::n endorsed by tae N IC . IJnanoroved rev's' ans to codes and standarc s wil' be reviewed on a case-by-case basis. In its letter cf May 13, 1992, ACRS agreed with the staff's position. Similarly, in its letter of May 5,1992, EPRI stated that the staff's position is consistent with the EPRI requirements documents. B. Electrical Distribution In SECY-91-078, " Chapter 11 of the Flectric Power Research Institute's (EPRI's) Requirements Document and Additional Evolutionary Light-Water Reactor (LWR) Certification Issues," dated March 25, 1991, the staff recommended that l the Commission approve its position that an evolutionary plant design should I include the following elements: l an alternative power source to the non-safety loads unless the design can demonstrate that the design margins will result in transients for a loss of non-safety power event that are no more severe than those associated with the turbine-trip-only event in current existing plant designs; and, i e at least one offsite circuit to each redundant safety division supplied directly from one of the offsite power sources with no intervening non-safety buses in such a manner that the offsite source can power the - safety buses upon a failure of any non-safety bus. l l l
i i i i 4 i In its SRM of August 15, 1991, the Commission approved the staff's positions. l In its letter of May 5,1992 EPRI indicated that this issue is not applicable to passive designs. However, the staff has not yet determined the
- applicability of this issue to the passive designs. This issue will be addressed in a separate Commission paper which will discuss the regulatory
- treatment of active nonsafety systems in passive plant designs.
j C. Seismic Hazard Curves and Design Parameters l To assess the seismic risk associated with an ALWR design, EPRI has proposed j the use of generic bounding seismic hazard curves for sites in the central and i eastern United States. EPRI proposes that these curves be used in the seismic j PRA. Current regulations do not require that a seismic PRA be performed to
- determine if a site is acceptable, and the staff does not intend to require i such an assessment.
To assess the EPRI ALWR seismic hazard bounding curve for rock sites, the staff compared the EPRI curve to results derived by Lawrence Livermore National Laboratories (LLNL). For this comparison, the staff used the historical earthquake method discussed in NUREG/CR-4885, " Seismic Hazard Characterization of the eastern United States: Comparative Evaluation of the LLNL and EPRI Studies," 1987. The staff also compared the EPRI bounding curve to hazard curves generated by EPRI using the historical method for the Seabrook site (see letter dated October 17,1991). The historical hazard curves below 0.lg reflect the past few hundred years of historical earthquake data. The historical hazard curves at higher accelerations are estimates based on the historical earthquake data. Both the LLNL and EPRI hazard curves, which were derived using the historical method, exceed the EPRI bounding curve at accelerations below about 0.19 Because the EPRI bounding curve is exceeded at low peak accelerations by the results based on historical earthquake data, the staff also questions the adequacy of the EPRI bounding curve at higher peak accelerations. hnard curves generated for the Seabrook Station Probabilistic Safety Assess-ment (1983) by the licensee also exceed the EPRI bounding hazard curve. The Seabrook SSE has a peak acceleration of 0.25g, whereas a higher SSE of 0.3g is proposed for ALWR sites. Based on the deterministic design basis of 0.3g, the EPRI-proposed criteria can be assumed to be suitable for the Seabrook site. However, based on the probabilistic assessment, the EPRI bounding hazard curve would underestimate the core damage frequency. Thus, the EPRI bounding hazard curve is non-conservative when compared to a licensee submittal. Similarly, the LLNL hazard curves used in the staff's reviews of seismic hazard are generally higher than the EPRI results for the same sites. Some LLNL hazard curves for sites in the Eastern United States ' discussed in NUREG/CR-5250, " Seismic Hazard Characterization of 69 Muclear Plant Sites East of the Rocky Mountains," 1989) exceed the EPRI bounding hazard curve.
During the staff's review of the ABWR, PRA results using both LLNL and EPRI hazard estimates were compared with results using the ABWR bounding seismic hazard curve. The ABWR bounding hazard curve was exceeded by the LLNL sean hazard curves for the Pilgrim, Seabrook, and Watts Bar sites. These three sites in the eastern United States were selected because of their relatively high seismic hazard. The staff used both LLNL and EPRI seismit hazard estimates to quantify core damage frequency. The PRA using the LLNL hazard curves predicted much higher core damage frequencies than the PRA using the EPRI hazard curve. However, the ABWR design was determined to be capable of resisting earthquakes significantly larger than an SSE of 0.3g. The evolutionary and passive ALWR designers have indicated that their applica-tions will be consistent with the EPRI criteria. However, based on review of historical seismicity and the LLNL hazard estimates, the staff concludes that the EPRI seismic hazard bounding curve is not sufficiently conservative. The staff is evaluating the seismicity and ground motion inputs used in the LLNL and EPRI studies to determine if the uncertainties in the curves can be reduced. To judge the seismic capability of the GE and ABB-CE designs for sites in the continental U.S., the staff used a deterministic process. On that basis, the staff concludes that, with few exceptions, most areas of the U.S. would be candidate sites for these designs. As part of the COL process, the applicant will have to demonstrate that the site-specific seismic parameters are within the bounding site parameters for the certified design. In its letter of May 5,1992, EPRI stated that they now specify a Seismic Margins Assessment (SMA) methodology, which plant designers can use to assess trie capability of advanced plants to shut down following a seismic event greater than an SSE. In addition, the PRA methodology may be used in evaluating a balanced seismic capability of standard designs, but calculation of risk as part of an overall core damage frequency determination is not required. The staff has reviewed this SMA methodology as part of its review of the EPRI requirements documents and should resolve any issues associated with this methodology through resolution of open items in the safety evaluation reports. The discussion on seismic hazard curves provided in this section is for infomation only. If a policy question is identified as a result of its review, the staff will infom the Commission of the issue at the earliest opportunity. D. Leak-Before-Break GDC 4 states, in part, that " dynamic effects associated with postulated pipe ruptures in nuclear power units may be excluded from the design basis when analyses reviewed and approved by the Commission demonstrate that the probability of fluid system piping rupture is extremely low under conditions consistent with the design basis for the piping."
Under the broad scope revision to GDC 4 (52 FR 41288, dated October 27,1987), the NRC allows the use of advanced technology to exclude from structural design consideration the dynamic effects of pipe ruptures in nuclear power plants. However, it must first be demonstrated that the probability of pipe rupture is extremely low under conditions consistent with the design bases for the piping. Demonstration of low probability pipe rupture requires a deterministic fracture mechanics analysis that evaluates the stability of postulated small, through-cll flaws in piping and the ability to detect leakage through the flaws long before the flaws could grow to unstable sizes. The concept underlying such analyses is referred to as " leak-before-break" (LBB). To date, the NRC staff has approved the LBB approach for currently operating and near-term operating licensed nuclear power plants based on a case-by-case review of plant-specific analyses. The NRC staff has approved the use of LBB for PWR primary coolant loop piping in all but five units in the United States. In addition, the use of LBB for pressurizer surge, accumulator, and residual heat removal piping have been approved for 11 units. In all cases, the LBB approvals have been granted for piping inside primary containment and for piping of at least 6 in. nominal diameter. The piping includes both austenitic and carbon steel material. However, all of the LBB-approved carbon steel piping has been clad with stainless steel material. To date, no BWRs have requested LBB approval. EPRI and GE have proposed to adopt the LBB approach for ALWRs when certain details of the piping design, material properties, and stress condition; are known. As discussed in SECY-89-013 " Design Requirements Related to the Evolutionary Advanced Light Water Reactors (ALWRs)," dated January 19, 1989, the staff will evaluate the acceptability of the use of LBB considerations in the ALWR designs when it can be justified. The staff has evaluated the EPRI and GE proposal for LBB application to ALWRs, as discussed below. LBB Acceptance Criteria The staff concludes that the analyses referred to in GOC 4 should be based on specific data, such as piping geometry, materials, and piping loads. The staff must review the LBB analyses for specific piping designs before the applicant can exclude the dynamic effects from the design basis. For ALWRs seeking design certification under 10 CFR Part 52, the analyses may be allowed to incorporate preliminary stress analysis results, provided that both upper-and lower-bounding limits are determined. Such limits would ensure that adequate margins are available for leakage, loads, and flaw sizes. A leakage margin of 10 is required to ensure that leakage from the postulated flaw size is detected when the pipe is subjected to normal operational loads. A load margin of 1.4 is required to ensure that leakage-size flaws are stable at normal plus accident loads (such as SSE and safety-relief valve discharge loads). A factor of 2 between the leakage-size flaw and the critical-size j flaw is required to ensure an adequate stability margin for the leaksge-size flaw. I
i l ;
- In addition, for ALWRs that seek NRC approval of LBB during the design j certification phase, certain information will be required for LBB analyses to
- establish through-wall flaw sizes and flaw stability. For through-wall flaw i sizes, a lower-bound, normal-operational stress limit must be established for
! dead weight, pressure, and thermal loadings. The mean or best-estimate l stress-strain curve should be used. For flaw stability, an upper-bound stress ! limit should be established for normal loadings plus safe SSE and suppression i pool hydrodynamic loadings. A lower-bound stress-strain curve for base metal i should be used, regardless of whether the weld or base metal is limiting. In j addition, a lower-bound toughness for the weld metal should be used. > A deterministic fracture mechanics evaluation accounting for material tough-1 ness is also required. An applicant may propose any fracture mechanics } evaluation method for NRC staff review. However, the applicant will have to { demonstrate the accuracy of the method by comparing it with other acceptable
- methods or with experimental data.
Using this approach, an initial set of bounding values and a preliminary LBB analysis can be established for ALWRs during the design certification phase. These bounding values and preliminary analyses can be verified when as-built and as-procured information becomes available during the COL phase. Before fuel-loading, the preliminary LBB analyses should be completely verified and based on actual material properties and final, as-built piping analysis as part of the inspections, tests, analyses, and acceptance criteria (ITAAC) associated with 10 CFR Part 52. LB8 Limitations Because of the dependency of th? LB8 analyses to accurately predict the flaw stability, the NRC has established certain limitations for excluding from the L88 approach piping that is likely to be susceptible to failure from various degradation mechanisms during service. A significant portion of the LBB review involves evaluating the susceptibility of the candidate piping in various degradation mechanisms to demonstrate that the candidate piping is not susceptible to failure from these degradation mechanisms. The NRC staff reviews the operating history and measures to prevent or mitigate these mechanisms. The LB8 approach cannot be applied to piping that can fail in service from such effects as water hammer, creep, erosion, corrosion, fatigue, thennal , stratification, and environmental conditions. The rationale is that these , degradation mechanisms challenge the assumptions in the LBB acceptance e criteria. For example, (1) water hammer may introduce excessive dynamic loads which are not accounted for in the LBB analyses, and (2) corrosion and fatigue may introduce flaws of a geometry that may not be bounded by the postulated througS-wall flaw in the LB8 analyses. Adhering to the " defense-in-depth" principle, piping susceptible to failure from these potential degradation mechanisms is excluded from LBB applications. i
Alternatively, features to mitigate the possibility of certain degradation mechanisms may be proposed to ensure that LBB assumptions are not invalidated. For example, LBB might be considered for carbon steel piping for which the effects of erosion and corrosion have been eliminated through the use of high chromium steels with proven resistance to erosion and corrosion or through the l use of carbon steel piping that is clad on the fluid-contacting surface with materials resistant to erosion and corrosion. A detailed discussion of the limitations and acceptance criteria used for LBB by the NRC staff is provided in NUREG-1061, Volume 3 " Evaluation of Potential for Pipe Breaks: Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," dated November 1984. Design Basis with LBB The broad scope rule introduced an acknowledged inconsistency into the design basis by excluding the dynamic effects of postulated pipe ruptures while ; retaining non-mechanistic pipe rupture for the containment, ECCS, and i environmental qualification (EQ) of safety-related electrical and mechanical equipment. The NRC staff subsequently clarified its intended treatment of the containment, ECCS, and EQ in the context of LB8 application in a request for l public comments on this issue that was published on April 6, 1988 I (53 FR 11311). ! Effects resulting from postulated pipe ruptures can generally be divided into local dynamic effects and global effects. Local dynamic effects are uniquely associated with a particular pipe rupture. These specific effects are not caused by a failure of any other source or even a postulated pipe rupture at a different location. Examples of local dynamic effects are pipe whip, jet impingement, missiles, local pressurization, pipe-break reaction forces, and decompression waves in the intact portions of that piping or communicating piping. Global effects of a pipe rupture need not be associated with a particular pipe rupture. Similar effects can be caused by failures of such sources as pump seals, leaking valve packings, flanged connections, bellows, manways, rupture disks, and ruptures of other piping. Examples of global effects are gross pressurization, temperatures, humidity, flooding, loss of fluid inventory, radiation, and chemical condition. For the A8WR, global effects also include suppression pool hydrodynamic loads (such as safety-relief valve discharges, pool-swell / fallback, condensation oscillation, and chugging loads). The suppression pool hydrodynamic loads caused by a main steam or feedwater pipe rupture might be excluded for the design of piping, equipment, and internal containment structures (other than those serving a containment function). Nonetheless, the possibility of such dynamic effects being caused by a reactor internal pump ejection, failures of flanged connections, and blowdowns from ruptured disks or squib-actuated valves have not been addressed at this time. The option does exist to establish a postulated pipe break of a high-energy line smaller than the main steam or feedwater line break to envelop the possible global dynamic effects described above. The
l
- designer would be required to submit this approach for NRC staff review and
- approval before use. Until then, the use of a postulated pipe rupture of a i main steam or feedwater line should be assumed for suppression pool
- hydrodynamic loads.
j The application of L88 technology eliminates the local dynamic effects of j postulated pipe ruptures from the design basis. Because the global effects l from the postulated pipe rupture provide a convenient and conservative design ) l envelope, the NRC staff will continue to require the consideration of global j effects for various aspects of the plant design, such as environmental
- qualification of equipment, design of containments, and design of subcompartment enclosures.
Recommendstions
- The revised GDC 4 of Appendix A to 10 CFR Part 50 permits elimination of the j dynamic effects of postulated high-energy pipe ruptures from the design basis
- of ALWRs using advanced fracture mechanics analyses. The limitations and i acceptance criteria for LBB applications in ALWRs are the same as those i established for currently operating nuclear power plants. Therefore. the i
4 staff recommends that the Commission anorove the anolication of the LBB anoroach to both evolutionary and cassive ALWRs seekina desian certification l i under 10 CFR Part 52. This anoroval should be limited to instances in which anorcoriate boundina limits are established usina oreliminary analysis results durina the desian certification chase and verified durina the COL ohase by nerformina the anoropriate ITAAC discussed herein. However, the specific details will need to be developed as the process is implemented during the first trial application. In its letter dated May 13, 1992, ACRS agreed with the staff's recommendation to extend the application of the LBB approach to both evolutionary and passive ALWR plants. Similarly, in its letter dated May 5,1992, EPRI agreed to extend the L88 approach to both evolutionary and passive plants and has since revised its URD to be consistent with the final staff's position stated above. The staff has not revised this position on LB8 since the draft Commission papers. E. Classification of Main Steamlines in Boiling Water Reactors (8WR) The main steamlines in BWR plants contain dual quick-closing main steam isolation valves (MSIVs). These valves isolate the reactor system in the event of a break in a steauline outside of the primary containment, a design basis LOCA, or other events requiring containment isolation. Although the MSIVs are designed to provide a leak-tight barrier, it is recognized that some leakage through the valves will occur. The current procedure for determining the acceptability of MSIV leakage
-involves calculating the dose in accordance with 10 CFR Part 100. This calculation is based on a conservative assumption that the leakage allowed by
l . the technical specification (normally 11.5 SCFH per valve) is released , directly into the environment. No credit is currently taken for the pressure integrity of the main steam piping and condenser. Because of recurring problems with excessive leakage of MSIVs, the staff developed guidance in RG 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants." This RG recommends the installation of a supplemental leakage control system (LCS) to ensure that the isolation function of the MSIVs complies with the specified limits. However, operating experience has shown that the LCS has required 1 substantial maintenance and resulted in substantial worker radiation exposure. 1 Additionally, the NRC has generic concerns with the effectiveness of the LCS l to perform its intended function under conditions of high MSIV leakage (Generic Issue C-8, Main Steam Line Valve Leakage Control Systems). These concerns led EPRI to propose an alternative approach to ensure that i doses associated with MSIV leakage would be acceptably low. EPRI identified ; this issue as a plant optimization subject. The resolution proposed by EPRI would allow higher leakage limits through the MSIVs, eliminate the safety-related leakage control system, and use an alternative MSIV leakage treatment j method that takes credit for the large volume and surface area in the main steam piping and condenser hotwell to plate-out the fission products following core damage. In this way, the main steam piping, main steam drain and bypass l lines, and condenser are used to mitigate the consequences of an accident. Appendix A to 10 CFR Part 100 requires that SSCs necessary to ensure the capability to mitigate the consequences of accidents remain functional during and after a SSE. These components are classified as safety-related and seismic Category 1. In addition, Appendix B to 10 CFR Part 50 establishes QA , requirements for safety-related, seismic Category I SSCs. j Section 3.2.2 of the SRP recommends that the main steamline from the outermost ! isolation valve up to, but not including, the turbine stop valve including 1 branch lines up to the first valve, be classified as Quality Group B (Safety l Class 2). RG 1.29 designates such piping as seismic Category I. ) l The staff concludes that the main steam piping from the outermost isolation ' valve up to the seismic interface restraint should confom to Appendix A to SRP Section 3.2.2 and RG 1.29, as should branch lines 24 inches in diameter and greater up to the first closed valve. The main steamline from the seismic interface restraint up to, but not including, the turbine stop valve should be classified as Quality Group 8, but may be classified as non-seismic Category I if it has been dynamically analyzed to demonstrate structural integrity under SSE loading conditions. However, all pertinent quality assurance requirements of Appendix 8 to 10 CFR Part 50 apply to this portion of the main steamline from the seismic interface restraint to the turbine stop valve. These requirements are needed to ensure that the quality of the piping material is commensurate with its importance to safety during both operational and
i - l ? i d accident conditions. In accordance with Position C.I.e of RG 1.29, the 2 turbine stop valve shall be designed to withstand the SSE and maintain its l integrity. 1 i The seismic interface restraint must provide a structural barrier between the i seismic Category I portion of the main steamline in the reactor building and i the non-seismic Category I portions of the main steamline in the turbine i building. The seismic interface restraint must be located inside the seismic ! Category I building. The classification of the main steamline in the turbine i building as non-seismic Category I is needed for consistency with the classi-fication of the turbine building. On this basis, the quality and safety requirements imposed on the main steamline from the outermost isolation valve . up to, but not including, the turbine stop valve are equivalent to the staff ! guidelines in Appendix A to SRP Section 3.2.2 and RG 1.29. ! The staff and EPRI agree that preventing gross structural failure of the j piping and hotwell would provide assurance that leakage from the MSIVs
- following a design-basis accident would not exceed the 10 CFR Part 100 guideline. In addition, this would ensure the integrity of the main steam
, drain lines and bypass piping from the first valve to the main condenser hotwell. The remaining issue is the classification of the main steam drain i lines and bypass piping between the first normally-closed valve and the i condenser hotwell as well as the hotwell itself. The staff crocoses that l neither the main steam drain and bvoass lines from the first valve un to the i condenser inlet. nor the cioina between the turbine ston valve and the turbine 4 inlet should be classified as safetv-related or as seismic Cateaory I. j Rather. these lines should be analyzed usino a dynamic seismic analysis to 4 demonstrate structural intearity under SSE loadina conditions. The turbine 1 ston. control. and bvoass valves and the main steam lines from the turbine I control valves no the turbine shall meet all of the cuality aroun and cuality ) assurance cuidel ines specified in SRP Section 3.2.2. Annendix A. Further. the staff crocosas that seismic analyses be nerformed to ensure that the condenser anchoraces~and the oicina inlet nozzle to the condenser are canable of maintainina their structural intearity durina and after the SSE. The dose analysis considers that the condenser is open to the atmosphere. Thus, it is only necessary to ensure that gross structural failure of the condenser will not occur. Similarly, it is only necessary to ensure that failure of non-safety-related SSC resulting from a seismic event will not cause failure of the main steam piping, main steam drain and bypass lines, or condenser. In its letter of May 13, 1992, the ACRS stated that it agrees with the staff's recommendation for the main steamline classification for both evolutionary and passive BWRs. In a letter dated May 5,1992, EPRI stated that it agrees with the staff's recommendations and proposed several clarifications, which the staff has incorporated into the above discussion. The staff concludes that the above-described approach to resolve the BWR main steamline issue for both evolutionary and passive ALWRs provides reasonable
- l
.i i i assurance that the main steam piping from the outermost isolation valve up to ! the turbine stop valve, the main steam drain and bypass lines up to the
- condenser, and the main condenser will retain their pressure and structural 1
integrity during and following a SSE. The staff recommends that the Commission aoorove the above-described aooroach to resolve the main steamline classification for both evolutionary and passive ALWRs. . F. Tornado Design Basis The current NRC regulatory position with regard to design-basis tornados is contained in two documents written in 1974: WASH-1300, " Technical Basis for Interim Regional Tornado Criteria," and RG 1.76 " Design Basis Tornado for Nuclear Power Plants." According to WASH-1300, the probability of occurrence of a tornado,that exceeds the design basis tornado (DBT) should be on the order of 10' per year for each nuclear power plant. The regulatory guide delineates maximum wind speeds of 386 to 597 kilometers per hour (km/hr) (240 to 360 miles per hour (mph)) depending on the regions. As a result of EPRI's earlier efforts on this EPRI-proposed plant optimization subject, the NRC reevaluated the regulatory positions in RG 1.76 using the considerable quantity of tornado data which has become available since the regulatory guide was developed. The reevaluation is discussed in , NUREG/CR-4661 (PNL-9697), " Tornado Climatology of the Contiguous United ' States," dated May 1986. At the heart of this study is the tornado data tape prepared by the National Severe Storm Forecast Center (NSSFC), which contains data for the 30 year period from 1954 through 1983. This tape contains data for the approximately 30,000 tornados that occurred during the period. The y staff determined that the tornado strike probabilities rangg from near 10 per year for much of the western United States to about 10' per year in the central United States. Based on discussions between the contractor and the staff, wind speed values associated with a tornado having a mean recur-rence interval of 10't per year were estimated to be about 322 km/hr (200 mph) for states west of the Rocky Mountains, and 482 km/hr (300 mph) for states east of the Rocky Mountains. In its letter of December 6,1991. EPRI proposed that the design-basis tornado to be used in the design of evolutionary ALWRs be based on a maximum tornado windspeedof482km/hr(300 mph)7 and a tornado strike probability derived from a recurrence interval of 10 per year. During a meeting with the staff on January 30, 1992, EPRI indicated that it would delete the reference to the tornado recurrence interval from the requirements document. The evolutionary and passive ALWR designers have indicated that their applications will be consistent with the EPRI requirements document. The tornado design-basis requirements have been used to establish struc-tural requirements (such as minimum concrete wall thicknesses) to protect nuclear plant safety-rolated SSC against effects not explicitly addressed in review guidance (such as RG or the SRP). Specifically, the staff has routinely reviewed and evaluated aviation crashes (involving general aviation
light aircraft), nearby explosions, and explosion debris or missiles, taking into account the tornado protection requirements. The staff's acceptance of EPRI's proposal will necessitate a concurrent review and evaluation of their effect on the protection criteria for some external impact hazards, such as general aviation or nearby explosions. Therefore, external impact hazards will be reviewed on a site-specific basis. Based on the updated tornado data and the analysis provided in NUREG/CR-4661, the staff concludes that it is acceptable to reduce the tornado design-basis ) wind speeds to 322 km/hr (200 mph) for states west of the Rocky Mountains, and ' 482 km/hr (300 mph) for states east of the Rocky Mountains. Therefore. the staff recommends that the Commission anorove the oosition that a maximum tornado wind soeed of 482 km/hr (300 mohl be used in the desian-basis tornada emoloved in the desion of evolutionary and passive ALWRs. The COL applicant will have to demonstrate that a design capable of withstanding a 482 km/hr (300 mph) tornado will also be sufficient to withstand other site-specific hazards. In its letter of May 13, 1992, ACRS agreed with the staff that the best , available data should be used to establish the tornado design basis. However, 1 ACRS noted the need to account for other potential loads that may previously have been subsumed within the tornado design-basis. The staff's position in the final policy paper has not changed, and the COL applicant will be required to demonstrate that the tornado design envelopes site-specific external impact hazards. In its letter of May 5,1992, EPRI agreed with the staff's position to use a , 482 km/hr (300 mph) maximum tornado wind speed and to consider other site- l specific hazards in the COL or early site permit. The staff expects that the use of these criteria will not preclude siting the ALWR plant designs on most sites in the United States. However, should an actual site hazard exceed the design envelope in a certain area, the COL applicant would have the option of perfoming a site specific analysis to verify that the design is still acceptable for that site. G. Containment Bypass The phenomenon of containment bypass is associated with either the failure of the containment system to channel fission product releases through the suppression pool, or the failure of passive containment cooling system heat exchanger tubes in the large pools of water outside the containment. The fundamental characteristic of a BWR pressure-suppression containment is that steam released from the reactor coolant system will be condensed (thereby limiting the pressure increase in the containment) and scrubbed of radionuclides in a pool of water (the suppression pool). This is accomplished by directing the steam from the reactor coolant system to the suppression pool through a vent system. However, leakage paths could exist (between the drywell and the wetwell airspace) that could allow steam to bypass the suppression pool and pressurize (or over-pressurize) the containment.
~
Potential sources of steam bypass include vacuum relief valve leakage, cracks in the drywell structure, and penetrations through the drywell structure. Therefore, the staff concludes that vendors should make reasonable efforts to minimize the possibility of bypass leakage and should account, in their containment designs, for a certain amount of bypass leakage. In addition, for a containment design that uses an external heat exchanger, the potential exists for containment bypass from a leak in the heat exchanger. High temperatures associated with severe accidents or core debris carried from the reactor vessel could threaten the integrity of the heat exchanger tubes and, therefore, provide a pathway for the release of fission products. Containment sprays in the drywell or wetwell would reduce the effect of suppression pool bypass leakage on containment performance. These systems spray water into the containment and lower its temperature and pressure. They also scrub the containment atmosphere of fission products and mitigate the effects of bypass on fission product distribution. In view of the contribution they can make to accident management, the staff is evaluating the need for containment spray systems for all ALWR designs. The GE ABWR and the ABB-CE System 80+ evolutionary designs have containment spray systems and, therefore, this issue is resolved for evolutionary designs. In its letter of May 5,1992, EPRI stated that they believed three separate topics were included in this issue. These topics were (1) interfacing system LOCA; (2) bypass of the suppression pool (BWR); and (3) failure of heat exchanger tubes in the passive containment cooling system (BWR). EPRI also indicated that their requirements documents adequately address these issues. The staff will address the issue of containment bypass and whether passive designs should contain a containment spray system in a separate Commission paper which will discuss resolution of issues related to source term. H. Containment Leak Rate Testing EPRI proposed that the maximum interval between Type C leakage rate tests should be 30 months, rather that the 24-month interval currently required by Appendix J to 10 CFR Part 50. This new maximum interval would apply to both evolutionary and passive plant designs. EPRI proposed this modification to allow some margin between the nominal 24-month refueling interval and the Type C test interval to ensure that plant shutdowns will not be required solely to perform Type C tests. Other issues (such as air lock testing and Type C leak testing methods) have also been raised, but have not been forwarded to the Commission as policy questions. In parallel with the staff's review of this ALWR issue, .the staff has developed proposed changes to Appendix J to 10 CFR Part 50 for all reactors. These changes were transmitted to the Commission in SECY-91-348, " Issuance of Final Revision to Appendix J to 10 CFR 50, and Related Final RG 1.JGX t (MS021-5)." This document proposes modification of the regulation that would i allow the increased interval and addresses other issues raised by EPRI. l SECY-91-348 also presents the staff's justification for these modifications.
This proposed rulemaking was withdrawn and is currently being reexamined under the Commission-approved prograe " Elimination of Regairements Marginal to Safety." The staff expects to complete this re-exataination and advise the Commission on its final decision about the rulemaki,ig course of action in December 1993. In addition, the staff has extended (by as much as 1 year) the time interval for performing Type C leakage rate testing on currently operating plants on a ; case-by-case basis. ' In its letter of May 13, 1992, ACRS identified no significant safety penalty caused by this change to the maximum interval between Type C leakage rate tests and agreed with the proposed staff position. In its letter of May 5, 1992,'EPRI recommended that the Commission approve the staff's position. Therefore. the staff recommends that the Commission avorove the cosition that. until the rule chanae croceedinas for Anoendix J of 10 CFR part 50 are completed. the maximum interval between Tvoe C leakaae rate tests for both evolutionary and nassive olant desians should be 30 months. rather than the 24-month maximum interval currently reautred in Anoendix J to 10 CFR part 50. I. Post-Accident Sampling System 10 CFR 50.34(f)(2)(viii) requires the designer to provide the capability to promptly obtain and analyze samples from the reactor coolant system and conteinment that may contain TID-14844 source term radioactive materials without radiation exposures to any individual exceeding 5 rem to the whole body or 75 rem to the extremities. Materials to be analyzed and quantified include certain radionuclides that are indicators of the degree of core damage (such as noble gases, iodines and cesiums, and non-volatile isotopes), hydrogen in the containment atmosphere, dissolved gases, chloride, and boron concentrations. RG 1.97 and NUREG-0737, " Clarification of TMI Action Plan Requirements," provide guidance regarding the design of the post-accident sampling system (PASS) used to implement 10 CFR 50.34(f)(2)(viii). EPRI has proposed deviation from several aspects of the PASS design requirements, as discussed below. EPRI has identified this issue as a plant optimization subject. Elinination of the Hydmgen Analysis of Containment Atmosphere Samples EPRI has stated that the hydrogen analysis of the containment atmosphere can be accomplished by the safety-grade containment hydrogen monitor required by 10 CFR 50.34(f)(2)(xvii) and Item II.F.1 of NUREG-0737 " Clarification of TMI Action Plan Requirements." The staff concludes that the safety-grade instrumentation provides adequate capability for monitoring post-accident
! l 4 l ) i hydrogen, and this is acceptable justification for an ALWR vendor to use in requesting this deviation. Because this exemption has previously been granted on currently operating plants, the staff does not consider this request to be a policy matter. In its letter dated May 13, 1992, ACRS agreed that elimination of the hydrogen analysis of containment atmosphere samples is appropriate, given that the safety grade hydrogen monitoring instrumentation will be installed. Elininstion of Olssolved Gas and Chloride Analyses of Reactor Coolant Sanples l EPRI considers the analyses of the reactor coolant for dissolved gas and chloride to be unnecessary because venting will remove gases accumulated in the reactor vessel (mainly hydrogen), and because prompt depressurization and cooling will minimize corrosion resulting from the presence of chloride and oxygen. Additionally, the amount of dissolved hydrogen in the reactor coolant can be detemined based upon the hydrogen concentration measured in the containment atmosphere. 10 CFR 50.34(f)(2)(viii) and Item II.B.3 of NUREG-0737 specify that the PASS should have the capability to analyze dissolved gases and chloride. This requirement was formulated before reactor vessels were required to have vents. With vented reactor vessels, the infomation on dissolved gas concentration became less important, l The staff concludes, however, that in PWRs even with vented reactor vessel l some postulated accident sequences can occur in which the reactor coolant i system is intact at reduced pressure, and heat is removed by subcooled decay heat removal (as in the TMI-2 accident). For these cases, it will not be possible to evaluate concentrations of the dissolved gases in reactor coolant by measuring their concentrations in the containment. For PWRs exposed to ' these conditions, infomation on the amounts of dissolved hydrogen in the reactor coolant is an important factor in evaluating post-accident conditions existing in the reactor vessel. The presence of hydrogen can affect flow of coolant in the core. Therefore, the staff concludes that for PWRs the requirement for PASS sampling 24 hours after the accident would be adequate to help ensure long-tem decay heat removal. In BWRs whenever core uncovering is suspected, the reactor vessel is depressurized to within approximately the pressure within the wetwell and the drywell. As a result of this decrease in pressure, dissolved gases will partially pass out of the water phase into the gas phase and under these conditicas the concentration of dissolved gases in the reactor coolant would have little meaning. During accidents in which only a small amount of cladding damage has occurred and the reactor vessel has not been depressurized, pressurized reactor water samples may be obtained from the process sampling system. Information on chloride and oxygen concentrations in the reactor coolant, although helpful in ensuring that proper steps are taken to minimize corrosion of reactor components, constitutes a secondary
r j . T I f , consideration since these samples could be taken at a low pressure. l Therefore, for BWRs, there will be no need for taking pressurized samples. 4 ! In its letter of May 5, 1992, EPRI stated that sampling of the reactor coolant ; i for dissolved gases and chlorides is not needed because ALWR coolant 1 l depressurization systems are highly reliable. Further, EPRI indicated that l i significant requirements associated with prevention of severe accidents make ! the need for PASS exceedingly small, especially in view of the limited
- usefulness to the operators of the information obtained from these measurements. In addition, even if a need for this infomation develops, EPRI l maintained that low levels of fuel failures predicted in ALWRs would permit
- operators to use routine sampling equip m t.
In its letter to ACRS dated June 12, 1992, the staff provided the justift-cation for retaining this samplir.g equips.ent for PWRs. The need for the ! dissolved gas sampling stems from the possibility of partially mitigated l severe accidents which do not involve early reactor depressurization, as was demonstrated in the THI-2 accident. In addition, there is a concern that these reactors may have a problem in maintaining reliable natural circulation and decay heat removal in the presence of non-condensible gases that would 1 evolve during depressurization. This concern is especially important in ! passive PWRs where decay heat removal systems are highly dependent on natural l circulation. Also, in these reactors, non-safety systems would need to be l used to perform the final cooldown depressurization. These actions would ! require that the operator fully appreciate the consequences of depressurizing l the plant and possibly introducing non-condensible gases that may interfere l with the successful termination of the event. l Determination of chloride concentrations, although helpful in ensuring that plant personnel take appropriate actions to minimize the likelihood of accelerated primary system corrosion following the accident, is a secondary consideration since long term samples could likely be taken at a low pressure. 1 It does not constitute, therefore, mandatory requirement of the PASS. l Therefore. the staff recommends that the Commission nonrove the position that I nost-accident smaalina systems for evolutionary and passive ALWRs of the s.ressurized water reactor tvoe be reautred to have the canability to analyze c issolved cases and chloride. in accordance with the reouirements of 10 CFR
$0.3Cff)f2)fv111) and Item III.B.3 of NUREG-0737. The time for takina these s ==a' es can be extended to 24 hours followina the accident. For evolutionary and passive ALWRs of the boilina water reactor tvoe. there would be no need for the oost-accident samalina system to analyze dissolved cases.
l Relantion in the Time Nequinnent for San \pling Activity Measurenents EPRI states in the Passive Requirements Document that if boron solution has been added to pemit plant shut down, reactor water samples can be taken for boron analyses starting eight hours after the end of power operation. EPRI also states that the samples for activity measurements will not be required i until the accident recovery phase. ' II
i ) i l Item II.B.3 of NUREG-0737 specifies that the PASS should have the capability i to obtain coolant and containment atmosphere sampling results within three hours following the accident. The purpose of this requirement is to ensure the capability to draw samples while the accident is still in progress, ' because analyses of the samples can provide insights for accident mitigation measures. EPRI has committed that the neutron flux monitoring instrumentation , will comply with the Category I critaria of RG 1.97. Therefore, this i instrumentstion will have fully 4 qualified, redundant channels that monitor j over the power range of 10 percent to full power. Based on this commitment, ! the staff concurs with EPRI's assertion that sampling for t,oron concentration i measurements will not be required for the first 8 hours after an accident. 1 i By contrast, samples for activity measurements are required to evaluate the
- condition of the core. During the accident management phase, this information i will be provided by the containment high-range area radiation monitor, the I
d containment hydrogen monitor, the reactor vessel water level indicator (for BWRs), and the core exit thermocouples (for PWRs). These data will be l sufficient to meet the needs of the plant operators for the first 24 hours j after an accident. The need for PASS activity measurements will ari!,e during 1 the accident recovery phase when the degree of core damage and general plant i contamination will have to be evaluated. Barea on this .1ustificatio.1, the staff concludes that the requested time extensf a for sampling activity measurements to 24 hours after an accident '.s acceptable, i Therefore. the staff recommends that the Commission anorove the deviation from )
~
the reautrements of Item II.B.3 of NUREG-0737 with reoard to reouirements f.gr ! samolino reactor coolant for boron concentration and activity measurements using the oost-accident samolino system in evolutionary and passive ALWRs. _ i The modified reouirement would reouire the canability to take boron concentration s== ales and activity measun. nts 8 hours and 24 hours. resoec-tively. followino the accident. 1 { J. Level of Detail 1
- i In its SRM of February 15, 1991, concerning SECY-90-377, " Requirements for ;
Design Certification Under 10 CFR Part 52," the Commission provided guidance l 1 regarding the level of detail of information required to determine the I adequacy of design certification applications under 10 CFR Part 52. Although i
- this issue is applicable to all design certification applications, the staff l
- has been reviewing the ABWR as the lead plant in resolving this issue. l j The staff identified several areas in the ABWR application where additional i information is needed in order to resolve safety concerns. The design detail
! resulting from the resolution of all of the staff's safety concerns will i constitute the level of detail needed to support design certification in j accortlance with the SRM on SECY-90-377. 4 i l i i a
l
- I I l
. i l
s ! The staff has informed the Commission of the progress of its efforts to i resolve the level of detail issue and has requested Comissio, guidance when appropriate. The following Commission papers have addressed issues related to
- level of detail-
! l l
- SECY-91-178, " Inspections, Test, Analyses, and Acceptance Criteria (ITAAC) for Design Certification and Combined Licenses;" l l
SECY-92-053, "Use of Design Acceptance Criteria During 10 CFR Part 52 , Design Certification Reviews;" , ! l i
- SECY-92-196, " Development of Design Acceptance Criteria (DAC) for the j Advanced Boiling Water Reactor (ABWR);"
4 ( SECY-92-214, " Development of Inspections, Test, Analyses, and Acceptance i j Criteria (ITAAC) for Design Certification;" and, ) l ? SECY-92-287, " Form and Content for a Design Certification Rule." a . , j The staff concludes that the level of detail issue is applicable to all design certification applications, but expects to resolve this issue in the context-J ! of the ABWR review. The discussion provided in this section is for ! J information only, and is provided to identify a complete list of issues ! i applicable to the passive designs. The staff is seeking no further guidance l
! from the Commissior,en this issue. '
i K. Prototyping } In SECY-91-074, " Prototype Decisions for Advanced Reactor Designs," the staff ;
- discussed the process it will use to assess the need for a prototype or other ,
1 demonstration facility for the advanced reactor designs. The staff stated '
! that it will follow the procedure outlined in SECY-91-074 to identify the ! various types of testing, up to and including a prototype facility, that may j be needed to demonstrate that the advanced reactor designs are sufficiently i nature to be certified.
Because the need for prototype testing is a design-specific issue, it cannot be resolved during the EPRI review. The staff has evaluated the submittals of i evolutionary ALWRs to assess the need for prototype testing and has not l identified any areas that may require such testing. l As discussed in SECV-91-273, " Review of the Vendor's Test Programs to Support l the Design Certification of Passive Light-Water Reactors," the need for separate effects and scaled integral testing for passive designs is under consideration. The staff it currently reviewing vendor test programs for the Westinghouse AP600 and the G5 SBWR and working closely with the vendors to j' resolve concerns. Since the issuance of SECY-91-273, the staff has forwarded to the Commission several Commission papers discussing integral system testing requirements and proposed NRC-sponsored confirmatory testing for the AP600 and e SBWR designs. l l
~
1 : i l The prototyping discussion provided in this section is for information only,- j and is provided to identify a complete list of issues applicable to the , passive designs. If a policy question is identified as a result of its l review, the staff will inform the Commission of the issue at the earliest , i opportunity. ! j L. ITAAC ITAAC are required for certified designs in accordance with Subpart B of l
- 10 CFR Part 52. Licensees that reference a certified design will implement !
[ the related ITAAC. The Nuclear Management and Resources Council (NUMARC) has ) 4 designated GE as the industry lead for developing ITAAC on their ABWR ! application. The staff is working with GE and NUMARC to develop ITAAC for
- design certification and has kept the Commission apprised of the status of j ITAAC. The following ITAAC-related papers have been sent to the Commission
i !
- SECY-91-178, "In:;pections Tests, Analyses, and Acceptance Criteria ,
} (ITAAC) for Design Certifications and Combined Licenses," June 12, 1991; 4
- SECY-91-210, " Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Requirements for Design Review and Issuance of a Final Design Approval," July 16, 1991; SECY-92-053, "Use of Design Acceptance Criteria (DAC) During 10 CFR Part 52 Design certification Reviews," February 19, 1992;
- SECY-92-196, " Development of Design Acceptance Criteria (DAC) for the Advanced Boiling Water Reactor (ABWR)," May 28, 1992;
- SECY-92-214. " Development of Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications," June 11, 1992; SECY-92-299, " Development of Design Acceptance Criteria (DAC) for the Advanced Boiling Water Reactor (ABWR) in the Areas of Instrumentation and Controls (I&C) and Control Room Design," August 27, 1992; ard,
- SECY-92-327, " Reviews of Inspections, Test, Analyses, and Acceptance Criteria (ITAAC) for the GE Advanced Boiling Water Reactor (ABWR),"
September 22, 1992. These papers discuss a wide range of issues related to ITAAC including how non-traditional issues discussed in this paper such as severe accidents and insights from the PRA are included in ITAAC. The discussion provided in this section is for information only, and fs provided to identify a complete list of issues applicable to the passive designs. The staff will continue to interact with the industry on this matter. Should additional policy questions be raised as a result of the staff't, review, the staff will inform the Commission at the earliest opportur,ity.
1 . t } ! M. Reliability Assurance Program In SECY-89-013, the staff stated that a program ca? led the reliability assurance program (RAP) would be required for design certification to ensure
, that the design reliability of safety significant SSC is maintained over the l life of a plant. In November 1988, the ALWR vendors and EPRI were informed j
4 that the staff was considering this matter. I The ALWR RAP would apply to those plant SSCs that are risk-significant (or ! significant contributors to plant safety) as quantified by the design certification PRA. The purposes of the RAP tre (1) to ensure that an ALWR is designed, constructed, and operated in a manner that is consistent with the i design assumptions for these risk-significant SSCs, (2) to prevent the reliability of these risk-significant SSCs from degrading during plant operations, (3) to minimize the frequency of transients that challenge ALWR i SSCs, and (4) to help ensure that these SSCs function reliably when
- challenged.
The staff views the RAP for ALWRs as a two-stage program. The first stage applies to the design phase of the plant life cycle, and is referred to as the design reliability assurance program (D-RAP). The second stage applies to the construction and operations phases of the plant life cycle and is referred to as the operational reliability assurance program (0-RAP). An applicant for design certification shall be required to establish the framework for the RAP (e.g., scope, purpose, objective, and essential elements of an effective RAP) and shall implement those portions of the D-RAP that apply to design g certification. A COL applicant would augment the design certification D-RAP with site-specific design infomation and would implement the balance of the D-RAP, including input to the procurement process. The COL applicant would also establish and implement the 0-RAP. The 0-RAP can be thought of as an umbrella program that would integrate aspects of existing programs (e.g., maintenance, surveillance testing, inservice inspection, inservice testing, and QA) to achieve its objective. The 0-RAP would apply to the construction and operation phases of plant life. i To ensure regulatory coherence between RAP and the maintenance rule (10 CFR 50.65), perfomance goals for risk-significant SSCs would be established under the 0-RAP based on input from the D-RAP. As such, performance and condition monitoring requirements for maintaining the reliability of risk-significant SSCs would be established. In addition, 0-RAP would provide a feedback mechanism for periodically re-evaluating risk significance based on actual equipment, train, or system performance. The majority of the 0-RAP would be based on the requirements of 10 CFR 50.65. The staff has completed its review of the AAP in the EPRI URD for the evolutionary ALWR, as documented in the final safety evaluation report (FSER) for Chapter 1, Section 6. The staff has also documented the results of its. review of the GE ABWR D-RAP in the dnft final safety evaluation report (DFSER) for Chapter 17.3. The staff has had the benefit of discussions with
; the ACRS regarding the form and content of the ALWR RAP. In its letter dated
l i . J i October 15, 1992, concerning proposed guidance for implementation of the maintenance rule (10 CFR 50.65), the ACRS noted the similarity between the
- maintenance rule, the license renewal rule, and the RAP. The ACRS stated that l consistent staff guidance is needed on the elements of an acceptable program j that will satisfy these three sets of requirements. The staff has
- incorporated the ACRS comments in developing its position on the RAP.
2 The following section summarizes the position the staff has taken in performing its review of D-RAP submittals contained in the design certification applications and the EPRI URD for the evolutionary and passive ALWRs. ! Interin Staff Position on RAP l t i For design certification, the staff position is that a high level commitment l to a RAP applicable to design certification (D-RAP) should be required as a ' non-system generic Tier 1 requirement with no associated ITAAC. The details of the D-RAP, including the conceptual framework, program structure, and l[ essential elements, should be provided in the SSAR. The SSAR for the D-RAP should also (1) identify and prioritize a list of risk-significant SSCs based !. on the design certification PRA and other sources, (2) ensure that the
- vendor's design organization determines that significant design assumptions, i such as equipment reliability and unavailability, are realistic and '
i achievable, (3) provide input to the procurement process for obtaining
- equipment that satisfies the design reliability assumptions, and (4) provide i
these design assumptions as input to the COL for consideration in the operational reliability assurance program (0-RAP). A COL applicant would
- augment the design certification D-RAP with site-specific design information
- and would implement the balance of the D-RAP, including input to the
] procurement process. The COL applicant would also establish and implement the 1 0-RAP. The staff will review the COL applicant's D-RAP and 0-RAP as part of l the COL application. t ; The staff position en 0-RAP will be provided in a future Commit;sion paper. In 1 developing its position on 0-RAP, the staff will ensure that regulatory l coherence exists between 0-RAP and the requirements of the maintenance rule I and the license renewal rule. l The staff's proposed resolution of the RAP for design certification will be provided to the Commission in a separate Commission paper which will discuss
- the regulatory treatment of active non-safety systems in passive plants. The j discussion provided in this section is for information only and is provided to identify a complete list of issues applicable to evolutionary and passive i
designs. The staff will continue to interact with the industry on this , matter. Statld additional policy questions be raised as a result of the
; staff's reviu, the staff will inform the Commission at the earliest
,' opportunity. 4
l l . { t l N. Site-Specific Probabilistic Risk Assessments and Analysis of External Events In its " Policy Statement on Severe Accidents Regarding Future Designs and ; Existing Plants " issued on August 8,1985 (50 FR 32138), the Commissio1 stated that applicants for future evolutionary reactor plant design reviews should complete a PRA. The Commission also stated that applicants should consider improving the means to avoid or mitigate severe accident vulnera-bilities exposed by the PRA in order to help ensure the public health and safety. Further, the Commission stated that evolutionary plant vendors should ! use the PRA in considering a range and combination of alternatives that address unresolved and generic issues and to search for cost-effective means to reduce the risk associated with severe accidents. In the policy statement, the Commission directed the staff to review evolutionary ALWR designs to detemine the safety acceptability of the design, stressing deterministic engineering analysis and judgment, complimented by PRA. After issuing this policy statement, the Commission promulgated 10 CFR Part 52. Section 52.47 requires that an application for design certification contain a design- i specific PRA. l In GL 88-20, " Individual Plant Examinations for Severe Accident Vulnerabil-ities - 10 CFR 50.54(f)," and its supplements, the staff stated that tonstruction permit holders and power reactor licensees should consider the ' safety implications of both internal and external events. Such consideration ! should involve performing separate individual plant examinations (IPEs) and i individual plant examinations for external events (IPEEE). PRAs and IPEs that ' have evaluated both internal and external events generally estimate the risks from external events to be the same order of magnitude as internal events. Therefore, the staff concludes that the design-specific PRAs required in 10 CFR 52.47 should include an assessment of both internal and external events. I Lessons from past risk-based studies indicate that fire, internal floods, and seismic events can be important potential contributors to core damage. However, the estimates of core damage frequencies for fire and seismic events j continue to include considerable uncertainty. Consequently, the staff concludes that fire and seismic events can best be evaluated using simplified i probabilistic methods and margins methods similar to those developed for existing plants, supported by insights from internal event PRAs (including ALWR design-specific PRAs). %e designer should use traditional probabilistic ] techniques to study internal floods. Fire events can be evaluated using simplified methods such as EPRI's Fire l Induced Vulnerability Evaluation (FIVE) methodology rather than full scala J PRAs. Ascribing to these methods, the designer would focus on the capacity of the design to withstand the effect of fire, using qualitative and quantitative methods rather than a strictly quantitative PRA fire analysis. The staff concludes that the plant designer can best determice the seismic capability of the plant through a combined approach that taket advantage of
l l l l 4 the strengths of both PRA and margins methods. This approach (based on an internal events PRA, its existing event and fault trees, and its random : failures and human errors) allows for a comprehensive and integrated treatment i of the plant's response to an earthquake. This approach should yield ! meaningful measures of a proposed design's seismic capability. l The major difference between a seismic PRA and the proposed PRA-based margins ! appNach is that the latter does not convolute fragility curves with hazard curves. Rather, the PRA-based margins approach measures the robustness of the i plant to withstand earthquakes of a given g-level. This method eliminates the need to deal with uncertainty in the seismic hazard curve for the site and ] i identifies potential design-specific seismic vulnerabilities. Understanding these vulnerabilities may be useful in developing the reliability assurance programs, identifying operator training requirements, and focusing on accident management capabilities. The staff will require each plant designer to perform a PRA-based margins analysis to identify the vulnerabilities of their design to seismic events. The plant designer should construct plant logic models covering the various systems that could be used to prevent core damage. Typically, this would be accomplished by modifying design-specific PRA models for internal events to ; include logic important in considering seismic failures. The models would not l include data from site-specific or generic seismic hazard curves. The ! designer would then determine all important accident sequences using the event i trees and fault trees (based on fragility data for each component for each { sequence). In addition, the designer would detemine the value of the minimum high confidence, low probability of failure (HCLPF) for the plant by deter- l mining the HCLPF values for the important SSCs for each accident sequences. i The HCLPF values calculated in this manner can be used to measure the plant's robustness and to provide an acceptable estimate of the earthquake ground , motion which the plant is expected to be able to survive without core damage. In general, the value of the plant HCLPF should be at least twice the design ground motion zero period acceleration. If not, the designer should perform a more detailed evaluation to identify any vulnerability against which the plant requires strengthened protection. HCLPF calculations also indicate which I components and systems limit the seismic capability of the plant. In its letter of May 8,1992, EPRI stated that they were unaware of suitable margins approaches to evaluate external events other than a seismic margins ' assessment. In its letter of August 21, 1992, EPRI provided additional comments regarding the staff's proposed requirements for analysis of certain internal and external events. EPRI concluded that their requirements for separation of redundant SSE by physical barriers designed to withstand full-height compartment flooding should make core damage events resulting from internal flooding very unlikely. In addition, EPRI concluded that the
>roposed recuirements for analyses of internal fire and seismic events would se addressed satisfactorily by their requirements documents, except for the provision of a seismic margin value twice the magnitude of the SSE. EPRI recommended that the seismic margin value be changed to 1.5 times the magnitude of the SSE to be consistent with operating plants. In addition.
~
} . i i i i EPRI noted that the as-built verification phase of the seismic margins evaluations can be based upon applying the appropriate margin to the site-i specific SSE. j The staff concludes that separated safety divisions do not necessarily preclude internal floods from being significant contributors to risk. The j staff notes that evolutionary ALWR designs that include physical separation j between safety divisions appear to respond better to internal fires and floods i than do traditional designs. However, although this physical separation { should reduce the expected frequency and severity of these events, it does not i
- preclude them from being important contributors to risk. For example, an '
4 internal flooding PRA study for an evolutionary design has already identified several design improvements that are needed to achieve the desired level of internal flood protection. The staff believes that the systematic evaluation required during the development of an internal flooding PRA is necessary to ensure that potential flood vulnerabilities are identified. The EPRI seismic margins methodology limits the search for success paths to two, and will not provide some important insights concerning human error and random failure. Therefore, this methodology does not appear adequate for evaluating evolutionary and passive ALWR designs. In addition, the EPRI i margint method is limited in its useful application for passive plant margins i analyses as is the NRC margins method used in the IPEEE, since these margins approaches are based on insights from PRAs of operating plants (PWRs and ' BWRs). The staff concludes that it would be inappropriate to rely exclusively on either of these methods to perform seismic margins analyses for passive ' designs. I In addition, the staff concludes that a well-designed plant should have a I plant HCLPF at least twice the magnitude of its SSE. Therefore, if the PRA- { based margins analysis only looks at accelerations up to 1.5 times the i magnitude of the SSE, the analysis could effectively screen out potential i design-specific seismic vulnerabilities. The staff would anticipate that l analysis of a fully developed seismic PRA will identify seismic avents as ; significant contributors to risk (perhaps the largest, given the significant 3 estimated reduction in internal event core-damage frequency in evolutionary l ALWx) . Therefore, it is important to fully understand potentially 1 significant seismic vulnerabilities and other seismic insights. This information would be captured by a PRA-based seismic margins analysis that considers sequence-level HCLPFs and fragilities for all sequences leading to core damage or containment failures up to approximately twice the magnitude of the SSE. Details of the specific site characteristics will likely not be available ! until the COL review stage. The staff intends to require a COL applicant to l perform a site-specific PRA that addresses all applicable site-specific hazards (such as river flooding, storm surge, tsunami, volcanism, or hurricanes). The staff will review the site-specific PRA to ensure that no vulnerabilities are introduced by siting the standardized plant at a location where external hazards could pose an unacceptable or unanticipated risk.-
i l ; In its letter of May 13, 1992, and in discussions with the staff, ACRS noted .i that the staff, an ALWR vendor, and a COL applicant may experience significant j obstacles if design vulnerabilities from site-specific external events are discovered at the COL review stage. ACRS requested more information on how the staff proposes to deal with unacceptable findings, resulting from the site-specific PRA, identified during the COL application review process. ) In SECY-92-287, " Form and Content for a Design Certification Rule," the staff ! proposed the appropriate processes to modify Tier 1 and Tier 2 design i certification information. For example, if a site-specific PRA identifies a
- serious generic design flaw that meets the
- adequate protection" threshold,
! the NRC can initiate rulemaking to amend the design certification rule. If l' the site-specific PRA identifies a site-specific design weakness, the COL applicant will have the option to request an exemption to the design l
certification rule to correct the deficiency. However, the staff will require that ALWR verfars perform bounding analyses of site-specific external events likely to be 7 challenge to a plant. When a site is chosen, its particular siting characteristics can then be compared to those of the bounding analyses in order to minimize the potential for the site-specific PRA to identify significant site-specific weaknesses for the standard design. Before certifying the design, the staff will evaluate fires, internal floods, and other external events that are not site dependent and will evaluate submitted bounding analyses for site-specific external events. l Therefore. the staff recomniends that the Commission anorove the position that l the analyses submitted in accordance with 10 CFR 52.47 should include an assessment of internal and external events. PRA insichts will be used to suonort a maroins-tvoe assessment of seismic events. A PRA-based seismic 1 marains analysis will consider secuence-level HCLPFs and fraailities for all l secuences leadino to core danace or containment failures un to anoroximately twice the maanitude of the SSE. Simplified crobabilistic methods. such as EPRI's FIVE methodoloov. will be used to evaluate fires. Traditional orobabilistic technioues should be used to evaluate internal floods. I Secondiv. the staff recommends that the Commission anorove the oosition that i ALWR vendors should perform boundino analyses of site-soecific external events i likely to be a challence to the plant (such as river floodina. store surae. l tsunami. volcanism. hiah winds. and hurricanes). When a site is chosen. its characteristics should be connared to those assumed in the boundin's analyses to ensure that the sito is envelooed. If P.he site is envelooed. t1e COL anolicant noed not neroors further PRA eva' untions for these external events. The COL ano' icant shou' d perform site-speciific PRA evaluations to address any site-soecifi c hazards for which a boundina analysis was not oer d orned or which are not enveloped by the boundino analyses to ensure that no vu' nerabilities due to sitino exist. The COL applicant should submit the results of the comparison of the bounding analyses to the site characteristics and any site-specific PRA information to the NRC at the COL review stage. For the GE ABWR and the ABB-CE System 80+
designs, the staff will work with the vendors to ensure a fundamental understanding of the potential vulnerabilities of their designs to external events. In order to maintain the design certification review progress, the staff will encourage, but not require, the vendors to complete these bounding analyses before issuance of a design certification rule. If the vendors choose not to perform the bounding analyses, they run a risk that site-specific hazards may limit acceptable sites for the design. For passive ALWR vendors who have submitted an application for design approval to the NRC, the staff will require that these bounding analyses be submitted and reviewed by the staff before issuance of final design approval. For subsequent design approval of evolutionary and passive ALWRs, the staff will require that the applicant submit bounding analyses for external events with the application. At the COL stage, the staff will review the site-specific characteristics to ensure that events enveloped by bounding analyses at the design stage or evaluated by a site-specific PRA have been properly addressed. The staff plans to conduct walk-down inspections to confirm that design commitments have been met. O. Severe Accident Mitigation Design Alternatives As discussed in SECY-91-229, " Severe Accident Mitigation Design Alternatives for Certified Standard Designs," dated July 31, 1991, the National Environmental Policy Act (NEPA), Section 102(C)(iii), requires, in part, that ! l
...all agencies of the Federal Government shall...(C) include in every recommendation or raport on proposals for legislation and other major Federal actions significantly affecting the quality of the human environment, a detailed statement by the responsible official on...(111) alternatives to the proposed action.
In Limerick Ecolooy Action v. NRC, 869 F.2d 719 (3rd Cir.1989), the U.S. Court of Appeals effectively required the NRC to include consideration of ' severe accident mitigation design alternatives (SAMDAs) in the environmental impact review performed as part of the operating license application for the Limerick Generation Station. A NEPA evaluation, in the form of an environmental impact statement that includes consideration of SAMDAs, is an essential element of an application for a COL under Subpart C of 10 CFR Part 52, for those applications that reference a design certified under Subpart 8. In SECY-91-229, the staff presented several options concerning the treatment of SAMDA issues as they related to the certification of standard plant designs. The staff recosmiended that the Commission approve the following recommendations:
- address SAMDAs for certified designs in a single 'rulemaking;
- approve the staff's approach for considering the costs and benefits of the review of SAMDAs for standard plant design certification; and, t
D
l I I approve the staff's proposal to advise applicants for a final design
- approval and design certification that they must assess SAMDAs and t provide the rationale supporting their decisions.
3 i In its SRM of October 25, 1991, the Commission approved the staff's I recommendations and requested that they be kept informed on the staff's ! progress in evaluating the SAMDAs for final design approval and design certification applications. Consistent with the third recommendation of SECY-91-229, the staff requested that ALWR vendors assess SAMDAs for their designs. This assessment and subsequent staff review is in addition to the safety consideration of severe accident issues discussed in this paper. The staff is currently addressing responses to this request in the context of the ABWR and System 80+ reviews. The discussion provided in this section is for infonnation only, and is provided to identify a complete list of issues applicable to the passive designs. The staff will continue to interact with the industry on this matter. Should additional policy questions be raised as a result of the staff's review, the staff will inform the Comission at the earliest opportunity. P. Generic Rulemaking Related to Design Certification In SECY-91-262, " Resolution of Selected Technical and Severe Accident Issues for Evolutionary Light-Water Reactor (LWR) Designs," dated August 16, 1991, the staff provided the Commission with recommendations regarding generic rulemaking related to design certification. The staff recomended that the Commission: -
- approve the staff's proposal to proceed with design-specific rulemaking thrnugh individual design certifications to resolve selected technical and severe accident issues for the A8WR and System 80+ designs; and,
. note the staff's intent to proceed with generic rulemaking, where appropriate for evolutionary and passive designs, as infomation becomes available from ongoing efforts on these issues, independent of the design review and certification processes.
The staff has not yet received Commission guidance on SECY-91-262. As discussed in SECY-91-262, the staff concludes that consideration of generic rulemaking in lieu of design-specific rulemaking is applicable to all final design approval and design certification applications. However, the design of the passive plants is not sufficiently developed at this time for the staff to determine whether generic rulemaking should be initiated for passive plant designs. Certain generic rulemaking activities related to the evaluation of source terms during postulated severe accidents are ongoing, and the results of these rulemakings may be used during design certification of the passive reactor designs. Currently, this work is focused on updating 10 CFR Part 100
2 I to separate siting criteria from reactor design criteria. The staff plans to incorporate the revised source-term criteria in 10 CFR Part 50. In addition, the staff plans to consider the incorporation of generic severe accident criteria. The discussion provided in this section is for information only, and is provided to identify a complete list of issues applicable to the passive designs. Should additional policy questions be raised as a result of the staff's review, the staff will inform the Commission at the earliest 3 opportunity. ) Q. Defense Against Common-Mode Failures in Digital Instrumentation and Control Systems Instrumentation and control (I&C) systems help ensure that the plant operates safely and reliably by monitoring, controlling, and protecting critical plant equipment and processes. The digital I&C systems for ALWRs differ signifi-cantly from the analog systems used in operating nuclear power plants. Specifically, digital I&C systems share more data transmission functions and shares more process equipment than their analog counterparts. Redundant trains of digital I&C systems may share data bases (software) and process equipment (hardware). Therefore, a hardware design error, software design error, or software programming error may result in a common-mode or common-cause failure of redundant equipment. The staff is concerned that the use of digital computer technology in I&C systems could result in safety- ' significant common-mode failures. The NRC staff developed these concerns in SECY-91-292, " Digital Computer Systems for Advanced Light Water Reactors." Some of the major points in that paper are suunarized as follows: Common mode failures could defeat the redundancy achieved by the hardware architectural structure, and could result in the loss of more than one echelon of defense-in-depth provided by the monitoring, control, reactor protection, and engineered safety functions performed by the digital I&C systems.
- The two principal factors for defense against common-mode and common-cause failures are quality and diversity. Maintaining high quality will increase the reliability of both individual components and complete e systems. Diversity in assigned functions (for both equipment and human i activities) equipment, hardware, and software, can reduce the 4 probability that a common-mode failure will propagate. 1 i
- The staff intends to require some level of diversity, such as a reliable analog backup. 4 Current regulations applicable to analog I&C systems also apply to digital I&C systems. In addition, the staff has developed limited guidance for digital I&C systems in RG 1.152, " Criteria for Programmable Digital Computer Software Systems in Safety-Related Systems of Nuclear Power Plants." However, as
i j i ! discussed in SECY-91-292, there are currently no regulatory requirements that , adequately address the potential safety concerns associated with digital I&C
- systems.
l Quality and diversity are important defenses against common-mode failures. i However, there are no standards for certifying the design of digital I&C i systems for application in nuclear power plants. In Enclosure 2 of i SECY-91-292, the staff discussed regulatory requirements that it is l considering to help ensure defense against common-mode failures, including the following areas:
- assessment of diversity; engineering activities; e
- design implementation; and, e safety classification of I&C systems.
The staff has made significant progress in establishing regulatory pidance that could be used to assess diversity. With the support of LLNL the staff has performed a study of the GE ABWR design to assess the adequacy of its , defense-in-depth and diversity. This assessment was performed using the method described in NUREG-0493, "A Defense-in-Depth and Diversity Assessment i of the RESAR-414 Integrated Protection System," for each transient and accident evaluated in Chapter 15 of the ABWR safety analysis report. The ' staff is using the results of this assessment to help determine the additional ' diversity necessary to defend against postulated common-mode software and ; hardware failures in the ABWR. I EPRI discussed requirements for engineering activities and design implementa- j tion for digital I&C systems in Chapter 10 of the EPRI ALWR for both evolu-tionary and passive plants. In SECY-91-292, the staff discussed the role of the EPRI URD as providing a frame of reference for the development of acceptance criteria for the design to adequately satisfy the requirements in j the URD. The criteria needed to satisfy the requirements for engineering activities and design implementation would be developed by the staff using applicable national and international standards where available, or expert opinions where adequate standards have not been developed, during the review of specific ALWR designs. As discussed in SECY-91-292, the staff is continuing to develop safety classification criteria for I&C systems in ALWR designs. The international technical community, through the International Electrotechnical Commission (IEC), recently issued an IEC standard, "The Classification of Instrumentation and Control Systems Important to Safety for Nuclear Power Plants." EPRI proposed certain classification positions in its "ALWR Position Paper for Passive System Classification and Requirements," submitted by a letter dated March 19, 1992. The staff is considering both of these documents before reaching a final position on safety classification criteria for I&C systems in
i f. i l passive ALWR designs. The safety classification of digital I&C systems relates to diversity through the defense-in-depth assessment by crediting i systems that have previously been classified as non-safety systems. Recently, increased attention has been given to detailed assessments of the ! integrity of software applied to safety-critical functions. These assessments ] have covered a broad range of applications, including computer-based medical treatment facilities, computer-based fly-by-wire aircraft control systems, and
- nuclear power plant protection systems. The staff found a consensus among computer science and software engineering experts that such safety-critical
- applications should be backed-up by some system not based on software. The
- experts based this opinion on the facts that the quantitative estimate for the i reliability of I&C systems based on high-integrity software cannot yet easily l
- be determined. The type of this backup and the functions it should perform depend on the specific equipment and design features of the I&C system.
] i The EPRI ALWR requirements document places special emphasis on common-mode failures to ensure they are addressed in man-machine interface system (M-MIS) ; i designs. EPRI stated that the ALWR Program has recognized from the onset that ! 1 there is currently no accepted standard to accurately quantify software ! reliability. To offset this concern, the ALWR Program has emphasized the need { for software quality and for a defense-in-depth approach to ensure the
- integrity of I&C functions including requirements for a backup hardwired manual actuation capability for system-level actuation of safety functions.
) As previously discussed, the staff has established potential regulatory
; guidance to ensure adequate diversity for digital I&C system applications. As
, a result of its review, the staff proposed an approach, in the draft Commission paper dated June 25, 1992, for assessing the defenses against common-mode failures in a design. The proposed approach also specified requirements for a backup system which is not based on software and which is used for system-level actuation of critical safety functions and displays of safety parameters. After carefully reviewing ACRS, industry, and vendor comments, the staff has developed a final position. The staff has concluded that analyses that demonstrate adequate, rather than equivalent, defense against the postulated cosmon-mode failures would be allowed in the diversity assessment required of i the applicant. The critical safety functions that require backup manual controls and displays would be specified. The staff would consider allowing 4 more flexibility in implementing the requirements for an independent set of i displays and controls. The necessary degree of flexibility depends on the i specific equipment and design features of the I&C system and will be evaluated for each design. The intent is to permit the use of diverse digital equipment that is not affected by the identified common-mode failures and to reduce complexity in the design. The staff will not require only analog equipment and will consider allowing simple digital equipment. i
} . , i 4 As a result of these changes, the staff revised the initial position proposed , 1 in the draft Commission paper. The staff recommends that the Commission l J approve the following revised staff position: '
- 1. The anolicant shall assess the defense-in-deoth and diversity of the l oronosed instrumentation and control system to demonstrate that vulnera-i bilities to common-mode failures have adeauately been addressed. The j staff considers software design errors to be credible common-mode j failures that must specifically be included in the evaluation. An 1 acceptable method of performing analyses is described in NUREG-0493, "A i Defense-In-Depth and Diversity Assessment of the RESAR-414 Integrated Protection System," March 1979. Other methods proposed by an applicant
- will be reviewed individually.
! 2. In oerformino the assessment. the vendor or aonlicant shall j analyze each costulated common-mode failure for each event that is j evaluated in the accident analysis section of the safety analysis i report (SAR). The vendor or anolicant shall demonstrate adeauate diversity within the desian for each of these events. For events 4 postulated in the plant SAR, an acceptable plant response should l not result in a non-coolable geometry of the core, violation of i the integrity of the primary coolant pressure boundary, or j violation of the integrity of the containment. 1
- 3. If a costulated common-mode failure could disable a safety function.
! then a diverse means. with a documented basis that the diverse means is unlikely to be subiect to the sgy coen-mode failure. shall be reouired to perform either the same function or a different function.
- The diverse or different function may be performed by a non-safety
- system if the system 'is of sufficient quality to perform the necessary
- function under the associated event conditions. Diverse digital or non-i digital systems are considered acceptable means. Manual actions frc=
4 the control room are acceptable if adequate time and information are t available to the operators. The amount and types of diversity may vary j among designs and will be evaluated individually. l
- 4. A set of safetv-arade disclavs and controls located in the main control
- room shall be provided for manual. system-level actuation of critical .
- safety functions and monitorina of narameters that suonort the safety i i functions. The disclavs and contro' s shal' be indeoendent and diverse
! from the safety computer system identified in items 1 and 3 above. The i specific set of equipment shall be evaluated individually, but shall be i sufficient to monitor the plant states and actuate systems required by j the control room operators to place the nuclear plant in a hot-shutdown i condition. In addition, the specific equipment should be intended to 1 control the following critical safety functions: reactivity control, I core heat removal, reactor coolant inventory, containment isolation, and containment integrity. / ) s The displays and controls shall be hardwired in the safety computer j system architecture to the lowest practical level. To achieve
! l l i l
! i i
I 5 system-level actuation at the lowest level in the safety computer system architecture, the controls may be hardwired either to analog components i ' or to simple, dedicated, and diverse software-based digital equipment that performs the system-level actuation logic. The safety parameter
- displays may include digital components exclusively dedicated to
' displays. This requirement would provide for an independent and diverse control logic for manual, system-level actuation of the safety function l that would be connected downstream of the lowest-level safety software-based component without affecting the hardware (interconnecting cables and interfaces) between the lowest-level electronic cabinets and the plant's electromechanical equipment. i Human engineering principles and criteria shall be applied to the i selection and design of the particular displays and controls. The i design of the displays and controls shall ensure that the human system ( ) interface shall be adequate to support the human performance require- j ments. i Hardwired, system-level controls and displays provide the plant operators with unambiguous information and control capabilities. These 3 controls and displays are required to be in the main control room to 1 i enable the operators to expeditiously mitigate the effects of the ! 1 postulated common-mode software failure of the digital safety I&C ' system. The control room would be the center of activities to safely i i cope with the event, which could also involve the initiation and imple- ' mentation of the plant emergency plan. The design of the plant should not require operators to leave the control room for such an event. For
- the longer term recovery operations, credit may be taken for actions from outside the main control room, when the emergency response organi-
- zation is fully briefed and in place to take such actions.
j R. Steam Generator Tube Ruptures ' i
; The staff has identified two distinct issues related to steam generator tube i ruptures (SGTRs). These issues, involving multiple ruptures specific to
- passive PWRs and containment bypass potential resulting from SGTRs, are l discussed below.
Multiple Steam Generator Tube Ruotures for Passive PWRs A design-basis accident involving SGTR in the current generation of pressurized water reactors is a rupture of one steam generator (SG) tube, with a rate of discharge of primary coolant through the SG tube break greater than the normal charging capacity of the reactor coolant inventory control system. SRP Section 15.6.3 requires the applicant to conduct an analysis for a single SGTR, but there is currently no requirement to perfom an analysis for multiple SGTRs. The staff is considering whether multiple SGTRs should be included in the plant design basis for advanced PWRs. a In NUREG-0844, "NRC Integrated Program for the Resolution of Unresolved Safety
l . i i 1 Issues A-3, A-4, and A-5 Regarding Steam Generator Tube Integrity," dated ' t September 1988, the staff estimated the probabilities of single and multiple [ tube ruptures. When the staff prepared these estimates in 1986, four single
- SGTRs had occurred in PWRs. All occurred in the United States and all the j affected plants were Westinghouse plants: Point Beach Unit 1 (February 1975);
- Surry Unit 2 (September 1976); Prairie Island Unit 1 (October 1979), and l R.E. Ginna (January 1982). Since that time, two more single SGTRs have s occurred in the U.S., one at North Anna Unit 1 (July 1987) and another at l NcGuire Unit 1 (Narch 1989).
I It i 10~pVREG-0844, the (RY). per reactor year staff estimated The staffthe frequency based of a single this estimate on SGTR to events the four be 1.5 x ! that occurred in approximately 300 " mature" reactor-years of Westinghouse i plant operation in the U.S. (" mature" RYs are accumulated after the firs't i 2 years of plant operation). ABB-CE and Babcock and Wilcox plants, which at that time had accumulated 77 and 66 mature RYs, respectively, without i experiencing any SGTRs, were assumed to have the same probability of SGTRs as i Westinghouse plants. In the same report, the staff estimated the probabi11ty
- of a multiple tube rupture event, using binomial statistics, as 1.6 x 10'3
} per RY. The staff based this estimate on a 50-percent level of confidence i (probability) for an event that had never occurred in the approximately 440 mature RYs accumulated among all U.S. PWRs at that time. ! Since the staff issued NUREG-0844, the total number of mature RYs of operation j for both Westinghouse PWRs and all U.S. PWRs has approximately doubled. i Westinghouse plants have now accumulated approximately 535 mature RYs, and all ? U.S. PWRs have accumulated about 827 mature RYs. Experience with Westinghouse plants (6SGTRsin535p'Ys)indicatesthatthefrequencyofasingleSGTRis approximately 1.1 x 10' per RY. A88-CE and Babcock and Wilcox p to have lower SGTR frequencies. With a failure rate of 1.1 x 10,} ants perappear RY, about 3 SGTRs should have statistically resulted in about 300 RYs of operating experience, and none have occurred. .The NRC has not received any report of an multiple SGTR in any U.S. or foreign plant. For consistency with the NUREG-0844 estimate of the probability of multiple SGTR events, the staff has , derived a new estimate based on a 50 percent confidence level for an event ! that has not occurred in approximately 827 RYs of U.S. PWR operation to date. 1 This estipated frequency for an multiple SGTR event is approximately ) 8.4 x 10 /RY. SGTRs are generally grouped into two categories: tho ' which occur as initiating events, and those which occur as a conseque:e of other events that increase the stress on the tubes. The probability estMates given above are ' for SGTRs that occur as initiating events. These events include random SGTRs caused by degradation of the tube over time, and SGTRs' caused by or associated with damage from foreign objects that may be present in the steam generator. Of the four SGTRs reported in NUREG-0844, two (Ginna and Prairie Island) are believed to have been caused at least in part by the impact of foreign objects on the steam generator tubes. The SG tubes in other plants have also leaked because of damage from foreign objects, although this leakage did not exceed the makeup capacity of reactor coolant inventory control systems. This issue
l l i is of concern in the context of determining the credibility of multiple SGTR
- events. While it would seem highly improbable that two randon SGTR failures
- would occur simultaneously (as indicated in NUREG-0844), damage or tube i failure caused by a foreign object could be a more likely initiator of an 1 multiple SGTR. In the Ginna event, the licensee enained the SG tubes after i the event and found that although only 1 SG tube hui ruptured, more than 20 l had been severely damaged.
The staff is reviewing the issue of whether to consider a single SGTR or an multiple SGTR as the design-basis event for the AP600. The staff is concerned that an AP600 plant could respond in substantially different ways to the two accidents, and that an multiple SGTR event could pose substantial challenges to the plant's passive safety systems. In dealing with an SGTR in a conventional plant, operators isolate the faulted SG and reduce the primary system pressure to help stop primary-to-secondary leakage. The operators use the safety-related pressure- and inventory-control systems in these plants (pressurizer spray, high-pressure safety injection) to carry out these procedures. While no multiple tube ruptures have occurred, leakage rate would likely increase with the number of tubes ruptured, and the operators would be required to act to mitigate the consequences of the event as quickly as possible. However, the basic procedures to be employed by plant operators in such an event would be similar to those used for single tube ruptures, and the plant conditions would probably be similar during the transient to those in a single SGTR. Analyses and tests of multiple tube rupture at the SEMISCALE facility have confirmed that the basic plant response is also similar to that for a single SGTR event. The AP600 plant includes no active safety-related inventory- or pressure-control systems. The core makeup tanks (CMTs) add high-pressure inventory by providing a gravity-driven injection of borated water. A natural circulation passive residual heat removal (RHR) system provides safety-related decay heat removal. The AP600 also uses an automatic depressurization system (ADS) to reduce the primary system pressure in the event of a LOCA and this ADS permits injection of. a large amount of low-pressure makeup water from the incontain-ment refueling water storage tank (IRWST). The first stage of the ADS is triggered upos rducing the CNT level to a predetermined point, with subsequent stages acttated as the CMT level decreases. The pressure of the primary system should be reduced to about that of the secondary system to inhibit primary-to-secondary leakage. However, primary system depressurization below that of the secondary system appears undesirable during an SGTR. Using the ADS will likely further lower the RCS pressure to such an ed.ent that unborated water could flow from the secondary side of the SGs back into the primary system. This could cause reactivity to increase in the core, with possible detrimental results. Westinghouse representatives assert that the AP600 has been designed with sufficient margin to the ADS initiation setpoint to allow at least 30 minutes of CNT injection after a single SGTR without triggering the ADS. Westinghouse
t i j ) i j representatives indicate that this should be sufficient time for the operators 3 to employ both safety-related and available non-safety-related systems to reduce RCS pressure, isolate the faulted SG, and teminate the event, j However, if an multiple SGTR occurs, with a substantially greater leakage rate
- of primary coolant, the AP600 might not be able to accommodate the accident j without actuating the ADS. The operators will have substantially less time to bring the event under control before the CMT level is reduced to the ADS i
setpoint. ADS actuation might then result in secondary-to-primary leakage of j unborated water. This water could flash to steam as it enters the RCS if the SG water is above the saturation temperature at the primary system pressure.
~
i' Since the passive safety systems of the AP600 rely on small differential pressures to circulate and inject emergency core coolant (ECC), introducing a large amount of steam into the RCS from flashing secondary water could disrupt 4 .or degrade ECC injection. Therefore, contrary to the response of current ! plants, the plant may respond to an multiple SGTR event in a manner i considerably different than that for a single SGTR. The consequences may also i differ significantly. , The designer could provide a number of methods to minimize the consequences of i multiple SGTRs, especially to retard or prevent secondary-to-primary leakage j or to lessen the amount of reactivity added as a result. These methods include (1) depressurizing the secondary system to maintain the RCS pressure at a value greater than the secondary pressure and prevent back leakage; ! (2) providing a system to borate the secondary water automatically if it leaks into the RCS from the steam generator; (3) or providing procedures that j inhibit ADS actuation if the primary-to-secondary barrier is breached. The i staff is not aware that Westinghouse is considering any of these SGTR mitiga-j tion methods for the AP600. l Design-basis accidents, such as large- and sapil-bregk LOCAs, have estimated i frequencies of occurrence on the order of 10' to 10' per RY. These events i generally provide the most rigorous test of the plant's safety systems. The j staffrecoggizes,hyver,thatthemultipleSGTRfrequencycouldbeinthe
- range of 10 to 10' per RY and that passive plant response for a multiple
, SGTR could significantly differ from that for a single tube rupture. This
- recognition has led the staff to conclude that rupture of more than a single j tube should be considered within the design basis of the plant.
l The staff is continuing to evaluate the appropriate number of ruptured SG l
- tubes that should be included in the design basis for the AP600. As a j minimum, the plant designer or applicant should analyze the single and i multiple SGTR events to detemine, to the extent possible, the quantitative j
! differences in the plant's response to these events. Therefore, as an interim l 1 step, the staff intends to require that Westinghouse analyze multiple ruptures 1 of two to five tubes in its AP600 safety analysis. The staff will report to
- the Commission when it determines the number of tube ruptures to be ,
, incorporated into the design basis of the AP600.
! l l i
1 j ! i EPRI stated that their URD require that future pressurized water reactors have substantially improved capability to handle SGTRs. EPRI also stated that i these requirements address material, design, and operation improvements to
- prevent SGTRs, a well as design features to improve the performance and j response of the plant after an SGTR. EPRI has also concluded that passive i plants are not unique with regard to multiple SGTRs, and accordingly, multiple
- SGTRs should not be included in the design basis for passive LWRs.
In its letter of September 16, the ACRS stated that it agreed with the staff's
- recommendation that Westinghouse should analyze the AP600 response to ruptures
- of up to five SG tubes and requested that the staff provide a "better technical basis" for estimating the frequency of multiple SGTRs.
! In its letter to ACRS of October 22, 1992, the staff noted that the calcula- l l tion for estimating the frequency of multiple SGTRs was not meant to represent a rigorous statistical analysis of multiple SGTR frequency. Rather, this ) l calculation was intended to show that the approximate frequency of multiple 1 SGTRs of a few tubes is on the same order of magnitude as that of other i limiting faults. The frequency calculation, coupled with the unique response I i of the passive safety systems to accidents involving loss of primary inven- , i tory, has led the staff to conclude that evaluation of multiple SGTRs up to l j about five tubes is warranted. The staff will use the vendor's analyses and i its own confirmatory analyses of these events to examine the response of the l AP600 to an multiple SGTR. The staff will subsequently recommend the extent i to which multiple SGTRs should be included within the design basis of the AP600. In letters of August 21, 1992, and September 17, 1992 EPRI and Westinghouse representatives, respectively, provided comments concerning the staff's position on multiple SGTRs. Westinghouse and EPRI believe that the design basis for the AP600 should remain a single tube rupture, and that multiple tube ruptures should be analyzed on a safety margin basis, using best-estimate techniques. Westinghouse and EPRI also indicated that they expect analyses of multiple SGTRs involving up to five tubes to show that the AP600 can respond to these events without actuating the automatic depressurization system. ' The staff does not share the views of Westinghouse and EPRI representatives. At this time, there is no experimental data on the performance of the AP600's passive safety systems that can be used to validate any models used for multiple SGTR analyses. The. staff therefore believes that there is sub- . . . stantial uncertainty in any such analyses. If the ADS were to actuate in the course of an multiple SGTR, the subsequent interactions could have a serious tapact on the ability of the passive systems to successfully terminate the accident. The staff expects that experimental data from planned vendor - testing programs will provide an adequate basis for evaluating the response of the AP600 to an multiple SGTR. The staff therefore maintains its stated position requiring analyses of multiple SGTRs up to five tubes for the AP600 and providing the results of l
l l 4 : l i
\
i these analyses in the application for design certification. The staff will i detemine the appropriate number of SGTRs that should be included in the AP600 ] design basis after evaluation of the submitted analyses. 4 J The staff recommends that the Commission anorove the cosition to reauire that . i analysis of multiole SGTRs involvina two to five SG tubes be included in the l 1 anolication for desion certification for cassive PWRs. The staff will { evaluate these analyses during the final design approval and design certifica- ! tion review process to help determine the number of SG tube ruptures that will j be incorporated into the passive PWR design basis. l Containment Bvoass Potential Resultino From SGTRs , ! The staff has identified an additional containment performance issue that has I ] not adequately been addressed. Specifically, a rupture of one or more SG l ) tubes could lead to a bypass of the containment. During a SGTR event, the SG i safety or relief valves could be actuated, discharging primary system radioactive inventory outside the containment. The staff concludes that the applicant for design certification should consider providing means to mitigate j this containment challenge. This issue applies to both evolutionary and
- passive PWR designs.
c , In its discussion on containment performance (Item I.J) in this Commission ,
- paper, the staff emphasized the importance of maintaining containment l
! integrity following a postulated severe accident. In its SRM on SECY-90-016, I } the Commission endorsed the staff's goal of reducing the probability for , conditional containment failure through the use of quantitative guidelines or l 1 alternative deterministic ob.jectives. The EPRI requirements document states : I that PWR containments should be designed to provide a leak-tight barrier to l l prevent uncontrolled release of radioactivity in the event of a postulated i (design-basis) a'ccident. Containment bypass due to SGTRs could violate ! containment integrity and hamper attainment of the severe accident goals j discussed in SECY-90-016. j The staff concludes that containment bypass resulting from SGTRs can be a
- significant challenge to containment integrity. Therefore, the staff
! concludes that the plant designer should consider design features that would i reduce or eliminate containment bypass leakage in such a scenario. The l following features could mitigate the releases associated with a tube rupture: i ! e a highly reliable (closed loop) steam generator shell-side hest removal system that relies on natural circulation and stored water sources; { I
- a system which returns some of the discharge from the steam generator j
relief valve back to the primary containment; or, l
- increased pressure capacity on the steam geaerator shell side with a j corresponding increase in the safety valve setpoints.
d
i i ! i 1 i l In its letter dated September 16, 1992, the ACRS Indicated that it agrees with the staff's recommendations to require that the applicant for design ! certification of a passive or evolutionary PWR assess design features i necessary to mitigate the amount of containment bypass leakage that could
- result from multiple SGTRs.
In a letter dated August 21, 1992, EPRI indicated that the requirements document requires that future PWRs have substantially improved capability relative to SGTRs, to minimize the potential for containment bypass and avoid repetition of past incidents, where SGTR resulted in continued lifting of steam safety valves. As such, the plant design shall be such that the complete and sudden rupture of one steam generator tube will not result in actuation of steam side safety valves. EPRI indicated that this policy issue relative to containment bypass is a design-specific issue, and relates to how a particular designer addresses the functional SGTR requirements already provided in the requirements document. In a letter dated September 17, 1992, Westinghouse indicated that the ALWR program developed in response to this issue includes Westinghouse input. The ALWR response to mitigate this issue includes the following elements: (1) both the evolutionary and passive PWR plants address SGTR containment bypass by features which significantly reduce the potential for core damage and for release directly to the atmosphere in SGTR accident sequences; (2) ALWR utilities are seeking design features in ALWR plants which could simplify operator response to SGTR; (3) evolutionary plants are designed to terminate SGTR by operator actions with a 30 minute grace period; and (4) passive plants have the same operator assisted capability but also include capability to mitigate SGTR without operator action. The sta*f recommends that the Commission ~ anorove the oosition to reouire that the anolicant for desian certification for a cassive or evolutionary PWt assess desian features to mitiente the == aunt of containment hvoass lea cace that could result from SG tube ruptures. The applicant or plant designer should consider the mitigation features that would likely be available following a postulated severe accident. The staff concludes that PWR designers should assess such features and address the desirability of this i mitigation function. The staff will review this issue when it performs the design certification review. S. PRA Beyond Design Certification , A plant-specific PRA is an excellent method for assessing overall plant safety and integrating plant systems and human interactions. Careful review of a PRA l can also revea important engineering evaluations, assumptions, and ' uncertainties. In the design and design review processes, PRA insights can be used to select among design options, strengthen the design against previously known vulnerabilities, characterize the design, and evaluate the balance in the design between severe accident prevention and mitigation. i
i l j At the COL stage, the COL applicant will be able to provide site-specific information and detailed design information that was not available during the 1 certification process. The COL applicant will be required to update the
; design-sacific PRA to reflect site-specific information before COL issuance.
During tie construction stage, the COL holder will also be able to consider as-built information. Experience has shown that subtle design interfaces involving support systems, systems interactions, or man-machine interfaces can j significantly affect the risk profile of a plant. l l The staff concludes that updated PRA insights, if properly evaluated and utilized, can strengthen programs and activities in areas such as training, i emergency operating procedure development, reliability assurance, maintenance, 4 and 10 CFR 50.59 evaluations. Therefore, the PRA should be revised to account ! for site-specific information, first-of-a-kind engineering developments, as-built (plant-specific) information refinements in the level of design detail, 4 plant operational experience, and design changes. The COL applicant or COL holder should update the PRA to ensure that new information or design changes do not introduce new vulnerabilities or diminish the overall capability of the design to prevent and mitigate severe accidents. As plant experience data accumulates, failure rates (taken from generic data bases) and human errors assumed in the design PRA should be updated and incorporated as appropriate, into Operational Reliability Assurance Programs. EPRI, Westinghouse, and NUMARC agree with the staff that a design-specific PRA has value and benefit, but they believe that the legal status of the PRA must be established under 10 CFR Part 52. Industry representatives expressed their desire to establish a common understanding of the legal and regulatory implications regarding the maintenance of the PRA. NUMARC, for example, stated that an ALWR design-specific PRA contains no unique or original design information that is not already reflected in associated SSAR Chapters. Accordingly, NUMARC believes that a design-specific PRA should be in neither Tier 1 nor Tier 2, but rather should be used as an analytical tool to assist the applicant and the NRC staff in evaluating the safety of the plant design. In its letter of September 16, 1992, ACRS agreed that it is worthwhile for a plant operctor to have an updated PRA. However, the ACRS was concerned about how the staff intends to use the updated PRA, how the staff thinks the licensee should use the updated PRA, and what should be required to update or keep the PRA current. The staff is considering additional regulatory requirements to address revising the design-specific PRA after it has been completed and will discuss this issue in a future Commission paper regarding the form and content of a I combined license. I e
j . I t { T. Control Room Annunciator (Alarm)' Reliability The annunciator system in a nuclear power plant provides a "first alert" to the control room operator of an abnormal stau in the plant, usually over the full spectrum of transients from the malfunctioning of a single piece of 4 equipment to the development of an abnormal state of one or more critical process parameters. The annunciator system also focuses the operator's i attention on the location and nature of the malfunction or disturbance. The extent to which this is achieved depends upon the design features of the annunciator system. ' Recent events at operating U.S. nuclear plants involving the loss of the plant i annunciator system have revealed that the power supplies of these systems are ! vulnerable tc single failures. At present, the NRC has no requirements ! specific to the annunciator system. The acceptance criteria and guidelines I for I&C systems (Appendix A to SRP, Section 7.1) developed from the GDC for j the I&C, control room, and protection and reactivity control systems, do not i specifically include the annunciator system. IEEE Standard 279, " Criteria for l Protection Systems for Nuclear Power Generating Stations," states that I protection systems design should provide the operator with information ' pertinent to its own status and to the generating station's safety. In a few special cases, specific alarms are required to comply with regulatory requirements because they are essential for the manual initiation of 1 protective actions. l When the operating U.S. plants were being designed, the international community generally observed the same requirements as those in the U.S. One of the few exceptions is International Electrotechnical Commission , Publication 231A, Supplement to Publication 231, " General Principles of l Nuclear Reactor Instrumentation," 196g. This publication gives specific ' requirements for the design of safety alanas but does not list their functional requirements. The international requirements changed in 1984 when the International Atomic Energy Agency published Safety Guide D8, " Safety-Related Instrumentation and Control Systems for Nuclear Power Plants." This Safety Guide discusses top-level requirements for those I&C systems, including the control room annunciator system, that perform functions important to safety but are not part of the traditional safety systems. Safety Guide D6 recosmiends a method for determining the relative importance to safety and the general principles for developing graded requirements for design features that determine the reliability and availability of these I&C 1 For the purposes of this paper, the annunciator system is considered to consist of sets of alarms (which may be displayed on tiles, video display units (VDUs), or other devices) and sound equipment; logic and processing support; and functions to enable operators to silence, acknowledge, reset and test alarms. l l l
systems. The staff discussed the need for such classification in some detail in SECY-91-292, " Digital Computer Systems for Advanced Light-Water Reactors," dated September 16, 1991. The EPRI URDs for both evolutionary and passive ALWR plants states the following requirements: The main control room (MCR) shall contain compact, redundant operator workstations with multiple display and control devices that provide organized, hierarchical access to alarms, displays, and controls. Each workstation shall have the full capability to perform MCR functions as well as support division of tasks between two operators. The display and control features shall be designed to satisfy existing regulations, for example: separation and independence requirements for Class IE circuits (IEEE Standard 384); criteria for protection systems (IEEE Standard 279); and requirements for manual initiation of protective actions at the systems level (Regulatory Guide 1.62 . defensive measures (e.g)., segmentation, fault tolerance, signalThe M-MIS desi validation, self-testing, error checking, and supervisory watchdog programs), as appropriate, to ensure that alarm, display, and control functions provided by the redundant workstations meet these standards. Thus, EPRI requires compact workstations with full ' capability to perfom control room. functions with fully organized alarms, displays, and controls. These workstations, including the alarms, are to be redundant and meet the requirements for independence and separation of Class IE and associated circuits described in IEEE 384, ' Criteria for Independence of Class IE Equipment and Circuits." This means that independence and separation must be provided between Class IE and non-Class IE circuits, even though the alam, display, and control devices are to be located in a workstation. The alarm system is considered nonsafety-related, and, therefore, the nonsafety-grade alarm circuits must be separated from interfacing Class IE circuits. The requirements for redundancy also apply to the power supplies associated with these work stations. These requirements fem a set of graded requirements for the alare, control, and leMication functions that implement the classification approach discussed in'SECY-91-292. The staff concludes that additional requirements for ALWR alarm systems are , necessary to minimize the problems experienced by operating nuclear power plants, such as the total loss of annunciators because of problems with their power supplies. The EPRI requirements for redundant workstations and displays that include the alarm functions are adequate for these stations. ' The staff recommendt that the Commission anorove the nosition that the alam ivstem for ALWRs should meet the ano' icable EPRI recuirements. as discussed above. for redundancy. Independence. and separation. In addition. alarms that
are provided for manually controlled actions for which no automatic control is provided and that are reouired for the safety systems to accomolish their safety functions. shall meet the acolicable recuirements for Class IE eouinment and circuits. III. Issues Limited to Passive Designs A. Regulatory Treatment of Nonsafety Systems in Passive Designs In contrast to both the current generation of LWR and the evolutionary ALWRs, the passive ALWR designs rely on safety systems that use the driving forces J buoyancy, gravity, and stored energy sources. . These passive systems supply safety-injection water, provide core and containment cooling, and perform other functions. There are no pumps in these passive safety systems, and all valves are powered by de electric power from batteries, are air-operated, or use check valves actuated by the pressure differential across the valve. EPRI and the passive reactor vendors contend that these designs do not. include safety-grade ac electric power. The passive ALWR designs also include nonsafety-grade active systems to provide defense-in-depth capabilities for reactor coolant makeup and decay heat removal. These systems serve as the first line of defense in the event of transients or plant upsets to reduce challenges to the passive systems. These active systems include: (1) the chemical and volume control system and control rod drive system, which provide reactor coolant makeup for the AP600 and SBWR, respectively; (2) the reactor shutdown cooling system and backup feedwater system for PWR decay heat removal, and the reactor water cleanup system for BWR decay heat removal; (3) the fuel pool cooling and cleanup system for spent fuel decay heat removal; and (4) the associated systems and structures that are needed to support these functions, including nonsafety standby diesel generators. In addition, the passive ALWR designs include nonsafety-grade active systems (such as the control room HVAC system) for mitigation of the radiological consequences of an accident. Many of these systems traditionally have been safety-grade systems, but in the passive plants, they are not designed to meet safety-grade criteria, and credit is not taken for them in the Chapter 15 licensing design-basis accident analyses. In SECY-90-406, " Quarterly Report on Emerging Technical Concerns," dated December 17, 1990, the staff identified the role of these nonsafety systems in the passive designs as an emerging technical issue. Associated with the new, passive design approach, the licensing design-basis analysis relies solely on the passive safety systems to demonstrate compliance with the acceptance criteria for various design-basis transients and acci-dents. However, uncertainties remain concerning the performance of the unique 1assive features and overall performance of core and containment heat removal acause of a lack of a proven operational perfomance history. For example, there are uncertainties about the performance of check valves in the passive safety systems, which operate at low differential pressures provided by natural circulation or gravity injection. These low pressures may not provide sufficient force' to fully open sticking check valves (that is, pumped ECCSs
l l i i t i are more likely to overcome stuck valves). These uncertainties enhance the igortance of the active nonsafety systems in providing the defense-in-depth , to prevent and mitigate accidents and core damage. Therefore, the staff's l review of the passive designs requires an evaluation of not only the passive
- safety systems, but also the functional capability and availability of the
- active nonsafety systems to provide significant defense-in-depth and accident and core damage prevention capability.
For active systems that perform defense-in-depth functions, the EPRI requirements document for passive designs specifies requirements concerning performance and systems and equipment design. These include radiation shielding requirements (to permit access following an accident), redundancy, availability of nonsafety-grade electric power, and protection against internal hazards. The requirements also address safety analysis and testing to demonstrate system capability to satisfy defense-in-depth considerations. EPRI does not currently provide specific requirements for the reliability of these systems. However, in response to staff questions, EPRI has indicated that it is evaluating specific reliability targets and other measures to provide confidence that the passive plants will meet performance requirements. These requirements will address both passive safety and active nonsafety systems. In addition, technical specification development is a subset of the overall regulatory treatment of the passive designs. The staff is evaluating the need to establish reliability-based technical specifications for passive designs. This evaluation will determine which systems and components (including certain nonsafety systems) require the imposition of technical specifications, and the parameters of the technical specifications. The Reliability Assurance Program is expected to strongly influence these technical specifications. Since the passive ALWR design philosophy departs from current licensing practices, new regulatory and review guidance is necessary so that the staff can appropriately review the AP600 and SBWR submittals. Significant decisions need to be made concerning the scope of staff review of the nonsafety systems and reliance on the passive systems. The staff will not require that the active systems meet all the safety-grade criteria, but there should be a high level of confidence that risk-significant active systems are designed in accordance with their performance / reliability missions to ensure their availability when needed. The staff has held several meetings with EPRI to determine steps needed to resolve the issue of regulatory treatment of active nonsafety systems, and define the scope of requirements and acceptance criteria to ensure that they have adequate capability and availability when required. In a meeting between NRC and the Utility Steering Committee on January 22,19g3, an agreement was reached for an overall process for determining the regulatory treatment of 4 nonsafety systems, and importance of passive systems and/or components for I meeting NRC Safety Goals and requirements. On February 23, 1993, EPRI submitted a document containing a draft proposed process. This document is under staff review. ' )
, . , , , w e--..
l ! l i l ! The staff is still evaluating this issue for t'a passive plant designs. The i discussion provided in this section is to inform the Counission of the current : i status of the issue. The staff's proposed resolution of this issue will be i provided to the Commission in a separate Commission paper, which will discuss j the regulatory treatment of nonsafety systems in passive designs. 1 . 1 l B. Definition of Passive Failure ' l Appendix A to 10 CFR Part 50 states that any applicant must design against ! l single failure of passive components in fluid systems important to safety, t where a single failure is defined as an occurrence which results in the loss { of a component's capability to perform its intended safety functions. Fluid and electric systems are considered to be designed against an assumed single i l failure if the system maintains its ability to perform its safety functions in I
- l. the event of a single failure of either any active component (assuming passive
{ components function properly) or a passive component (assuming active i componients function properly). However, the introduction to Appendix A to l 10 CFR Part 50 notes that the conditions under which a single failure of a ! passive component in a fluid system should be considered in designing the j system against a single failure are under development. ? l SECY-77-439, " Single Failure Criterion," describes how the staff was using the j single failure criteria in its reactor safety review process. As discussed in j that paper, an active failure in a fluid system means (1) the failure of a ! component which relies on mechanical movement to complete its intended function on demand, or (2) an unintended movement of the component. Examples include the failure of a motor- or air-operated valve to move or to assume its
- correct position on demand, spurious opening or closing of a motor- or air-j operated valve, or the failure of a pump to start or stop on demand. In some
- instances, such failures can be induced by operator error.
4 i A passive failure in a fluid system means a breach in the fluid pressure t boundary or a mechanical failure which adversely affects a flow path. Examples include the failure of a simple check valve to move to its correct position when required, the leakage of fluid from failed components (such as pipes and valves) particularly through a failed seal at a valve or pump, or line blockage. Motor-operated valves which have the source of power locked out are allowed to be treated as passive components. In past licensing reviews, the staff has been inconsistent in its treatment of passive failures in fluid systems. Specifically, the staff imposed a passive failure in addition to the initiating event but not in others. The staff has determined that, in most instances, the probability of most types of passive , failures in fluid systems is sufficiently small that they need not be assumed ! in addition to the initiating failure in application of the single failure l criterion to ensure the safety ,of a nuclear power plant. In particular, staff practice'has nomally been to treat check valves, except for containment isolation systems, as passive devices (rather than active devices) during transients or design-basis accidents. However, the staff is
l
- f
- considering redefining check valve failure as an active failure. This change 6
appears necessary because safety-related check valves in the passive designs will operate under different conditions (low flow and pressure without pump
- pressure to open valves) than current generation reactors and evolutionary i designs. In addition, they have increased safety significance to the i operation of the passive safety systems, and operating experience has shown j
that they have a lower reliability than originally anticipated. Redefining check valve failure in this manner would cause these valves to be evaluated in { a more stringent manner than that used in previous licensing reviews. 4 The staff is still evaluating this issue for the passive plant designs. The staff's proposed resolution of this issue will be provided to the Commission in a separate Commission paper, which will discuss the regulatory treatment of j active nonsafety systems in passive designs. i 4 C. SBWR Stability 1 1 i In BWRs, thermal-hydraulic instabilities can cause oscillations that can i ! result in violation of the minimum critical power ratio (MCPR) safety limits. ! l The staff has concluded that GE's analytical codes have been sufficiently ! validated to demonstrate the stability of the ABWR design. However, the codes !
- that Gl: is using have not yet been adequately validated for the passive BWR '
j design. As discussed in SECY-91-273, " Review of the Vendor's Test Programs to Support !
- the Design Certification of Passive Light-Water Reactors," the staff detemined that an early NRC assessment is needed. This assessment should 1 l
i address the vendor's analytical and experimental basis for demonstrating 4 nuclear / thermal-hydraulic stability. In addition, it should identify any tests or analyses that may be needed to support the staff's technical evaluations of the issue. The NRC staff and its consultant, Oak Ridge National Laboratories (0RNL), have reviewed the thermal-hydraulic stability
- characteristics of the SBWR bastd on preliminary design information provided by GE. This assessment included calculations with the LAPUR computer code j developed by NRC and ORNL. These calculations shwed that, while the system appears to be very stable under normal operating conditions, certain abnormal operating conditions might be reached under credible transient sequences.
These abnormal conditions can result in the onset of density-wave power and flow esc 111ations. In addition, a low-flow and low-power instability caused by a "geysering" effect between parallel channels hes been identifid as a concern during normal operating transients such as start-up and shutdown. On December 6,1991, the ' staff met with EPRI and GE to discuss the EPRI/GE response to the staff's conclusions. Specifically, the ttaff conclued that more extensive S8WR stability studies are needed and that codes which have been validated against thermal-hydraulic tests representative of the S8WR design (including the large open chimney) would be nr.eded to perform these studies. EPRI and GE informed the staff that the etieney design has been changed and that existing experiments are representative of the divided chimsey now employed. GE also indicated it will validate its codes for
. ~ . _ . . - - . . - - - - - - - . - . . - - . . - . - . . . . - . - - - - . - -
t l I l ! l l l ! ! density-wave instability studies against these experiments and will provide 3
- the results of this work for NRC review. The geysering instability is being j j ttudied using small-scale experiments performed by a Japanese partner to GE.
< ?he Japanese SAFAR code will be validated against these experiments and used l hr analytical prediction of stable operating boundaries. GE plans to ; i recommend start-up/ shutdown procedures similar to those used in the Dutch ! j Dodewaard reactor to avoid geysering instability. In addition, EPRI and GE l l believe that the SBWR is not vulnerable to a loop-type instability reported by 1 the Japanese; rather, EPRI and GE contend that this instability was charac- i j teristic of the experimental apparatus used. ! In SECY-92-339,
- Evaluation of the General Electric Company's (GE's) Test l Program to Support Design Certification for the Simplified Boiling Water Reactor (SBWR)," the staff noted that GE has modified the SBWR conceptual )
design. The staff also noted that GE has identified existing experimental l data, which they believe constitutes appropriate validation of codes to be ! used for stability studies. EPRI and GE have indicated agreement with the I staff that such studies will be needed to confirm the stability of the SBWR l during various transient scenarios (including ATWS). However, GE has not provided sufficient infomation to permit NRC evaluation of the applicability : and sufficiency of the foreign experiments they have identified for use during code validation. The vendor has agreed to make this infomation available to the NRC as soon as it obtains' permission from the foreign sources. Until these experiments can be reviewed by NRC, the potential need for additional l experiments to support stability evaluations for design certification remains open. l The staff considers this a technical issue and expects to resolve this issue ' with GE through its nomal review of the SBWR design certification application and through its review of GE's S8WR testing program. The staff will interact with the Commission if additional policy issues are identified during its review of the 58WR application or vendor S8WR testing program. D. Safe Shutdown Requirements GDC 34 of Appendix A to 10 CFR Part 50 requires that a residual heat removal system be provided to remove residual heat from the reactor core so that specified acceptable fuel design limits (SAFDL.s) are not exceeded. RG 1.139 and Branch Technical Position (8TP) 5-1 implement this requirement and set forth conditions for ceid shutdown g3.3 *C (200 *F) for a PWR and 100 *C (212 *F) for a 8WR) using only safety-grade systems within 36 hours. The RG presents the basis for this requirement, as follows: 1
...even though it may generally be considered safe to maintain a reactor in a hot standby condition for a long time, experience shows that there have been events that required eventual cooldown and long-term cooling until the reactor coolant system was cold enough to perfom inspection and repairs. It is therefore obvious l 4 that the ability to transfer heat from the reactor to the I environment after a shutdown is an important safety function for I i
j . . l I both PWRs and BWRs. Consequently, it is essential that a power ! plant have the capability to go from hot-standby to cold-shutdown conditions...under any accident conditions. i Because passive ALWR designs use passive heat removal systems for decay heat i removal, they are limited by the inherent ability of the passive heat removal processes. These designs cannot reduce the temperature of the reactor coolant
- system below the boiling point of water for the heat to be transferred to the
- in-containment refueling water storage tank of the AP600 or the isolation i'
condenser of the SBWR. Even though active shutdown cooling systems are available to bring the reactor to cold-shutdown or refueling conditions, these active RHR systems are not safety-grade and do not comply with the guidance of RG 1.139 or BTP 5-1. l EPRI states that it is not necessary for passive safety systems to be capable j ~ ef achieving cold shutdown. EPRI bases this contention on the belief that the l passive decay heat removal (DHR) systems have an inherently high long-term l reliability. The EPRI requirements document for passive plant designs states ! < that the passive ALWR designs will employ a redundant safety system for both ! l the hot-standby and long-term cooling modes. In addition, it defines safe 1 3 shutdown as 215.6 *C (420 *F). EPRI has indicated that it meets GDC 34 , requirements because redundant passive decay heat removal systems can operate l at full RCS pressure and place the reactor in the long-tem cooling mode . immediately after shutdown. Additionally, EPRI requires that operation of the l 1 plant in the long-term cooling mode be automatic, eliminating operator actions 1 to cool down the plant. Also, operation of the passive DHR system does not ! require any ac power or pumps. EPRI further states that the nonsafety systems that will take the plant to cold-shutdown conditions "...are highly reliable ] in their own right...and a failure in these systems would not prevent the j plant from achieving cold shetdown." l The staff is currently evaluating the EPRI position eith . respect to this issue to assess the acceptability of their proposed alternative approach for meeting GDC 34. The long-ters DHR capability of the proposed passive systems offers potential advantages over current active systems. However, the staff must resolve several issues before reaching a final position on this matter. These issues include reliability criteria for the nonsafety systems which have the capability to bring the plant to cold shutdown and the acceptability of 2I5.6 *C (420 *F) as a safe, long-tem state. The staff's proposed resolution of this sue will be provided to the Commission in a separate Commission paper, which will discuss the regulatory treatment of nonsafety systems in passive designs. E. Control Room Habitability GDC 19 of Appendix A to 10 CFR Part 50 states that adequate radiation protec-tion shall be previded to pemit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem phole body, or its equivalent to any part of the body, for the duration of the accident. In current plants, safety-grade filtered control
i-I i, - j i i 1 room HVAC systems with charcoal absorbers are used to ensure that radiation { doses to operators could be maintained within the GDC 19 criteria in the event i of an accident. 4 In SRP Section 6.4, the staff defined the acceptable operator dose criterion 1 in terms of specific whole body and organ doses (5 rem to the whole body, and i 30 rem each to the thyroid and skin). Recently, the NRC embraced the princi-i pal recommendations of Publication No. 26 of the International Commission on i Radiological Protection in the promulgation of a major revision of 10 CFR j Part 20. The adoption of these recommendations, which include use of the i effective dose equivalent, did not change the dose criteria for control room
- operators to conform with GDC 19.
J j Originally, EPRI proposed an expesure limit for control room operators of 1 5 rem whole body, 75 rem skin, and 300 rem thyroid. EPRI stated that each operator would be provided with individual breathing apparatus and protective 3 clothing, if required, to meet regulatory limits. The staff determined that , t EPRI's requirements for the thyroid and beta skin doses were not adequately justified. The staff indicated to EPRI that the long-tern use of a breathing j apparatus during design-basis vcidents has never been allowed. More , j importantly, the long-term use of a breathing apparatus is likely to degrade ' } operator performance during and following an accident. EPRI also stated that the control room would be designed to be maintained during a 72-hour period as the primary location from which personnel can
- safely operate in the event of an accident. It was and is the staff's l position that, depending upon the accident, the required duration may be much longer than 72 hours. GDC 19 states that " adequate radiation protection shall
- j. be provided to permit access to and occupancy of the control room under i accident conditions...for the duration of the accident." Consequently, the staff concluded that analyses of control room habitability should consider the
! duration of the accident (which may extend well beyond the EPRI-proposed j 72-hour period) as the design basis. l In order to resolve this matter, the staff proposed that EPRI and the vendors j provide a high level of assurance that the control room ventilation system , j will be available when needed. Because.the system may not need to meet all of i the safety-grade criteria, it may be appropriate to allow some credit for , nonsafety-grade ventilatica and filtration systems based on reliability 1 considerat<ons. The extent of this credit will be determined as part of the j staff's review of the regulatory treatment of nonsafety systems. It should be
- noted that unlike the case of core cooling, there is no passive safety-grade
] system for defense-in-depth of control room habitability. -
- In its letter of May 5,1992, EPRI proposed an alternative which would use a
- i safety-grade pressurization system capable of being recharged remotely after l 72 hours. In that enclosure, EPRI stated that Volume III of the Utility
; Requirements Document would be revised.to require: (1) a passive, safety- ! , grade control room pressurization system which would use bottled air to keep l operator doses within the limits of GDC 19 and SRP Section 6.4, Revision 2 for i
a
1 l l t l i the first 72 hours of the event, and (2) safety-grade connections for the pressurization system to allow use of offsite, portable air supplies if needed after 72 hours to minimize operator doses. The staff agrees with EPRI's i commitment to limit the operator doses to those specified in GDC 19 and SRP Section 6.4, Revision 2. However, the staff is still evaluating the proposal j to utilize a safety-grade pressurization system and has serious reservations ! concerning the feasibility and the capability of a pressurization system to ! maintain the control room habitability. l In its letter of August 17, 1992, ACRS indicated that they had discussed the i subject 1.f control room habitability with EPRI and the staff during a meeting j on June 4 and 5, 1992. At that meeting, the staff told ACRS that they were ! evaluating the EPRI proposal utilizing the safety-grade pressurization system. ! ACRS indicated that they had several comments regarding the design features of ! the passive control room pressurization system proposed by EPRI. ACRS stated { that the staff should take these comments into account. In performing its
- evaluation, and that ACRS may provide additional recommendations after the j staff has completed its evaluation.
The staff is currently reviewing the new severe accident source term proposed by EPRI in conjunction with the staff's technical update of the TID-14844 source term. The estimated potential radiological consequences to the passive plant control room operators during a severe accident will depend on the outcome of the forthcoming resolution of severe accident source ters. Specifically, this outcome will include chemical forms of fission products, release fractions, and release timing. In addition, the control room habitability assessment is further dependent upon the fission product removal processes inside, as well as outside, of the primary containment before it reaches the control rooer air intake and-the control building that houses the control room. Therefore, the staff is unable to complete its control room habitability assessment until issues concerning the source term and its behavior mechanism are satisfactorily resolved. The staff plans to present its proposed resolution of this issue in a separate Commission paper, which will discuss issues related to source term. F. Radionuclide Attenuation EPRI and the passive Al.WR designers rely on assumptions involving fission product removal inside containment by natural removal effects and holdup by the secondary building and piping systems. A containment spray system is not mandated by the EPRI requirements document for passive plant design . The staff is concerned about the uncertainty in quantifying the holdup phenomena in the auxiliary building and that use of the auxiliary building for holdup may require imposition of additional restrictions on the auxiliary building i during nomal operation, with which the licensee may have difficulty ! complying. i l
_ ~ __ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _ _ _ _ l - e j ! This issue also affects the control room habitability issue discussed in Section III.E of this report. The relationship arises because the industry indicates that fission products will be removed before they reach the control room air intake or the control building that houses the contr31 room. The staff is still evaluating this issue as well as the need for a containment spray system for the passive plant designs. The staff is also evaluating : whether credit for fission product attenuation in the main steamlines and condenser is appropriate for the passive BWR design. This question arises i because the main steamlines downstream of the main steam isolation valves and ! associated condenser are not designed to withstand the SSE, as defined in Section III.c of 10 CFR Part 100. The staff concludes that plateout of radioactive iodine on the main steam pipe and condenser surfaces following a severe accident can realistically provide significant dose mitigation. i Several technical references indicate that particulate and elec.tal iodines ! would be expected to deposit on surfaces with deposition rates varying with temperature, pressure, gas composition, surface material, and particulate size. The staff's proposed resolution of this issue will be presented in a separate l Commission paper, which will discuss source term related issues. ' G. Simplification of Offsite Emergency Planning EPRI has proposed to significantly simplify offsite emergency planning for passive designs because of EPRI's estimated low probability of core damage and, in the event of a core damage accident, the assurance of containment integrity and low offsite dose. EPRI's proposal would eliminate requirements for early notification of the public, detailed evacuation planning, and provisions for exercising the offsite plan. The onsite emergency plan and limited offsite actions would be retmed. EPRI has identified this matter as a plant optimization subject. During a meeting with the staff on January 30, 1992, EPRI proposed to work with the staff to define a process for addressing simplification of emergency ! planning. This simplification would include developing technical criteria and methods that, if met, would justify such action. It would also include defining the process for implementing this approach. The results of this effort would be used as input to a generic rulemaking proposal to be initiated ; by NUMARC.
~
The staff concludes that certain modifications to the emergency planning requirements of 10 CFR Part 50 and the siting criteria in 10 CFR Part 100 may be appropriate for the passive designs based on their unique characteristics. However, an agency detemination on these issues will requira evaluation of detailed design infomation. The staff concludes that the unique character-istics of these designs should be taken into account in detemining the extent of emergency planning required for the plume exposure pathway emergency plann-ing zone. Any decision on emergency planning requirements for the passive design should reflect a plant's ability to prevent the significant release of
^
) i l I radioactive material or to provide very long delay times before a release for i all but the most unlikely events. Before relaxing emergency planning require-ments, the staff will require a high degree of assurance that all potential containaient bypass accident sequences have a very low likelihood. The lack of information concerning source term and risk precludes further staff evaluation of the emergency preparedness requirements for the passive reactor designs at j this time. Moreover, the issue is cc::: plicated by the fact that the promul-t d gation of emergency planning requirements folkwing the TMI-2 accident was not premised on any specific assumptions about severe accident probability. Hence, as a policy matter, it mai be that even very low calculated probability l i values should not be considered a sufficient basis for changes to emergency 1 i planning requirements. s
- The staff will evaluate this issue for the passive plant designs when suffi- 1 cient supporting information becomes available. The staff plans to update the j status of this review in a separate Commission paper.
H. Role of the Passive Plant Control Room Operator In SECY-91-272, " Role of Personnel and Advanced Control Rooms in Future ! Nuclear Power Plants," dated August 27, 1991, the staff discussed the role of ! the operator in a passive plant control room. Specifically, operators in a t passive plant control room may use nonsafety-related systems and active
" investment protection" systems as the primary means to mitigate transients and accidents. Operators will use these systems, before safety-related passive systems are initiated, when responding to transients and accidents.
The design of safety-related systems in the passive plant differs
.significantly from the design of safety-related systems in current operating plants and in evolutionary plant designs.
To safely operate a passive plant, the operator must understand the operation of the " investment protection" systems and their interfaces with the safety-related passive systems. Passive plant operators will be required to perfom new functions and tasks unlike those for evolutionary plants. These new functions and tasks will be associated with the new operational philosophy noted above, the increase in automation, and the greater use of advanced technology in passive plant designs. These new functions and tasks will likely involve greater reliance on monitoring and decision-making rather than perfoming actions directed in procedures. Thus, the design process must carefully define the operator's role to ensure that it properly develops the man / machine interface design to facilitate these functions and tasks. EPRI stated that the ALWR Program has provided for " man-in-the-loop" testing during first-time engineering, as specified within Chapter 10 of the EPRI URD. EPRI also requires a full scope control room design simulator for this test-ing. To correct the problem of insufficient focus on the operator in previous designs, EPRI indicated that this requirement should adequately ensure that the human component in the man / machine interface system is explicitly included. However, EPRI maintains that the difference in the role of the operator in a passive plant control room is limited to the details and timing L -
1 4 i f 1 of the actions performed. The staff will consider the operator's role in its review of each passive plant applicant's control room design under 10 CFR j Part 52. l The staff concludes that an extensive man-in-the-loop test and evaluation program will be necessary for the passive plant control room designs. This i testing will address the extent of differences in the operator's role in a passive plant control room since it will simulate tasks necessary for maintaining plant safety following an event. Such testing would likely j require a fully functional integrated control room prototype to demonstrate l s that the passive designs properly consider the operator's role for ensuring j plant safety. ! In its letter of August 21, 1992, EPRI restated that the approach to operating
- passive designs is the same as for evolutionary or existing designs. The staff does not agree with this position. However, the EPRI ALWR Program has j
provided for man-in-the-loop testing and a full scope control room design ! simulator to ensure that the human component is explicitly considered and is , l acceptable. The staff has continued discussions with EPRI and passive plant ' vendors and believes this approach will resolve the differences in the position on operating philosophy. 4 In its letter of September 16, 1992, ACRS agreed with the staff that sufficient man-in-the-loop testing and evaluation should demonstrate that the , operator's functions and tasks are properly integrated into the man / machine d interface design. . Therefore. the staff recommends that the Cmission anorove the oosit. ion that sufficient man-in-the-loco testino and evahuation must be nerformed. In addition. a fully functional inteorated control room crototvoe is likely to be necessary for cassive olant control room desions to demonstrate that functions and tasks are croceriv inteorated into the man / machine interface desion. These requirements will be incorporated into the DAC. Each applicant may provide justification that a control room prototype of reduced scope is
, sufficient to ensure that functiont and tasks are properly integrated in the man / machine interface design.
4 1
ALWR ISSUES CROSS-REFERENCE MATRIX Cateaory Issue Title Commission PaDers I. SECY-90-016 A. Use of Physically Based Source 86-228 Issues Tere 88-203 89-013 89-153 89-228 89-341 90-016 90-307 90-329 90-341 90-353 92-127 Draft-l' B. ATWS 89-153 89-228 90-016 90-353 Draft-1 C. Mid-loop Operation 89-228 90-016 90-353 , Draft-1 D. Station Blackout 89-013 ! 89-153 l 89-228 - 90-016 l 90-329 ' 90-353 Draft-1 E. Fire Protection 89-013 89-228 I 90-016 l 90-353 Draft-1 1 l '" Draft-1" refers to the draft Commission paper, " Issues Pertaining to Evolutionary and Passive Light-Water Reactors and Their Relationship to
- . Current Regulatory Requirements," which was fonvarded to the Commission on February 20, 1992, and made available to the public on February 27, 1992.
ENCLOSURE 2 I
\ l 1
Cateoorv' Issue Title Commission Pacers 1 I. SECY-90-016 F. Intersystem LOCA 89-153 j Issues (cont.) 89-228 ! i 90-016 90-353 Draft-1 G. Hydrogen Control 89-013 l i 89-153 89-228 90-016 90-329 90-353 l Draft-1 l H. Core Debris Coolability 89-153 l 89-228 90-016 90-353 ! 92-092 l Draft-1 l t ', I. High-Pressure Core Melt 89-228 i Ejection 90-016 90-353 92-092 Draft-1
- J. Containment Performance 89-228 3 90-016 90-353 91-273 92-092 Draft-1 K. Dedicated Containment Vent 89-153 l Penetration 89-228 a
90-016 90-329 l 90-353 92-092 < Draft-1 L. Equipment Survivability 88-228 50-016 90-353 Draft-1 M. Elimination of OBE 89-013 90-016 90-329 90-353
i Cateaory Issue Title Commission Papers I. SECY-90-016 M. Elimination of OBE (cont.) 91-135 Issues (cont.) Draft-1 Draft-2 2 N. In-Service Testing of Pumps 89-228 and Valves 90-016 ; 90-353 i 91-273 ! Draft-1 I II. Other Evolu- A. Industry Codes and Standards 91-273 tionary and Draft-1 Passive De-sign Issues 8. Electrical Distribution 91-078 Draft-1 C. Seismic Hazard Curves and 91-13! Design Parameters Draft-1 D. Leak-Before-Break 89-013 Draft-1 E. Classification of Main Steam- Draft-1 lines in BWRs F. Tornado Design Basis Draft-1 G. Containment Bypass Draft-1 H. Containment Leak Rate Testing 89-013 89-228 91-273 Draft-1 I. Post-Accident Sampling System Draft-1 J. Level of Detail 90-241 90-377 , Draft-1 i K. Prototyping 91-074 91-273 Draft-1 1 1 1
** Draft-2" refers to the draft Commission paper, " Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light-Water Reactor Designs," which was forwanied to the Commission on June 25, 1992, and made available to the public on July 1,1992.
Cateaory Issue Title Commission Paoers II. Other Evolu- L. ITAAC 91-178 tionary and 91-210 Passive De- 92-053 sign Issues 92-196 (cont.) 92-214 92-287 92-294 Draft-1 M. Reliability Assurance Program 89-013 92-133 Draft-1 N. Site-Specific PRAs and 89-013 Analyses of External Events Draft-1 Draft-2
- 0. SANDAs 91-229 Draft-1 P. Generic Rulemaking Related to 91-262 Design Certification Draft-1 Q. Defense Against Common-Mode 91-292 i Failures in Digital I&C Draft-2 l Control Systems R. Multiple SG Tube Ruptures 92-133 Draft-2 S. PRA Beyond Design Certifica- Draft-2 tion T. Control Room Annunciator Draft-2 Reliability III. Issues A. Regulatory Treatment of Active 89-013 Limited to Nonsafety Systems 90-406 Passive 92-133 Design Draft-1 Draft-2
- 8. Definition of Passive Failure 77-439 Draft-1 C. Thermal-Hydraulic Stability 89-153 of the S8WR 91-273 92-339 Draft-1 i'
l A Cateaorv Issue Title Commission Papers III. Issues D. Safe Shutdown Requirements Draft-1 Limited to 92-131 , Passive ! Design E. Control Room Habitability 92-133 (Cont.) Draft-1 Draft-2 F. Radionuclide Attenuation 92-127 92-133 : Draft-1 ! 1
! G. Simplification of Offsite 88-203 1 Emergency Planning Draft-1 i 1 , H. Role of the Passive Plant 91-272 !
l Control Room Operator Draft-2 I i a i i.
COMMISSION PAPERS APPLICABLE TO ALWRs i SECY-77-439, " Single Failure Criterion," August 17, 1977. ! SECY-86-228, " Introduction of Realistic Source-Ters Estimates into Licensing,"
- August 6, 1986.
i SECY-88-147, " Integration Plan for Closure of Severe Accident Issues," May 25, 1988. i l, SECY-88-202, " Standardization of Advanced Reactor Designs," July 15, 1988. i SECY-88-203, " Key Licensing Issues Associated with DOE-Sponsored Advanced j Reactor Designs," July 15, 1988. SECY-89-012 " Staff Plans for Accident Management Regulatory and Research Programs," January 18, 1989. i SECY-89-013 " Design Requirements Related to the Evolutionary Advanced Light-Water Reactors (ALWRs)," January 19, 1989. SECY-89-153, " Severe Accident Design Features of the Advanced Boiling-Water Reactor (ABWR)," May 10, 1989. SECY-89-178, " Policy Statement Integration," June 9,1989. SECY-89-228, " Draft Safety Evaluation Report on Chapter 5 of the Advanced Light-Water Reactor Requirements Document," July 28, 1989. SECY-89-341, " Updated Light-Water Reactor (LWR) Source-Tern Methodology and Potential Regulatory Applications," November 6,1989. SECY-90-016, " Evolutionary Light-Water Reactor (LWR) Certification Issues and Their Relationship to Current Regulatory Requirements," January 12, 1990. SECY-90-065, " Evolutionary and Passive Advanced Light Water Reactor Resources and Schedules," March 7,1990. SECY-90-146, " Process, Schedule, and Resources for the Review of Evolutionary and Passive Advanced Light-Water Reactors," April 20, 1990. SECY-90-241, " Level of Detail Required for Design Certification Under Part 52," July 11.1990. SECY-90-307, " Impacts of Source-Ters Timing on NRC Regulatory Positions," August 30, 1990. SECY-90-313, " Status of Accident Management Program and Plans for Implementa-tion," September 5, 1990. 1 Enclosure 3 l
I i i SECY-90-329, " Comparison of the General Electric Advanced Boiling-Water , j Reactor (ABWR) Design and the Electric Power Research Institute's (EPRI's) i Advanced Light-Water Reactor (ALWR) Requirements Document," September 20, < l 1990. l f SECY-90-341, " Staff Study on Source-Ters Update and Decoupling Siting from l 1 Design," October 4, 1990. ' i SECY-90-353, " Licensing Review Basis for the Combustion Engineering, Inc. ! j System 80+ Evolutionary Light-Water Reactor," October 12, 1990. ' l l SECY-90-377, " Requirements for Design Certification Under 10 CFR Part 52," ; l November 8, 1990. ' d j SECY-90-406, " Quarterly Report on Emerging Technical Concerns," December 17, j 1990. 2 i SECY-91-074, " Prototype Decisions for Advanced Reactor Designs," March 19, 1991. l SECY-91-078, " Chapter 11 of the Electric Power Research Institute's (EPRI's) Requirements Document and Mditional Evolutionary Light-Water Reactor (LWR) 2 Certification Issues," March 25, 1991. ! SECY-91-135, " Conclusions of the Probabilistic Seismic Hazard Studies Con-3 ducted for Nuclear Power Plants in the Eastern United States," May 14, 1991. SECY-91-161, " Schedules for the Myanced Reactor Reviews and Regulatory Guidance Revisions," May 31, 1991. { SECY-91-178, " Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC)
- for Design certifications and Combined Licenses," June 12, 1991.
1 l SECY-91-210, " Inspections, Tests, Analyses, and Acceptance Criteria (ITAAC) Requirements for Design Review and Issuance of a Final Design Approval,"
~
July 16, 1991. l l SECY-91-229, " Severe Accident Mitigation Design Alternatives for Certified Standard Designs," July 31, 1991. l l SECY-91-239, " Preapplication Reviews of Myanced LWR Designs," August 5,1991. 2 SECY-91-262, " Resolution of Selected Technical and Severe Accident Issues for Evolutionary Light-Water Reactor (LWP.) Designs," August 16, 1991. l SECY-91-272, " Role of Personnel and Myanced Control Rooms in Future Nuclear Power Plants," August 27, 1991. I SECY-91-273, " Review of Vendors' Test rams to support the Design Certifi-cation of Passive Light Water Reactors," ust 27, 1991. SECY-91-292, " Digital Computer Systems for Myanced Light-Water Reactors," September 16, 1991.
l l ' SECY-91-348, " Issuance of Final Revision to Appendix J to 10 CFR 50, and Related Final Regulatory Guide 1.XXX (NS 021-5)," October 25, 1991. . SECY-92-030, " Integral System Testing Requirements for Westinghouse's AP600 Plant," January 27, 1992. j SECY-92-037, "Need for NRC-Sponsored Confirmatory Integral System Testing of 3 the Westinghouse AP600 Design," January 31, 1992. SECY-92-053, "Use of Design Acceptance Criteria During 10 CFR Part 52 Design j Certification Reviews," February 19, 1992. SECY-92-092, "The Containment Perfomance Goal, External Event Sequences, and i the Definition of Containment Failure for Advanced Light-Water Reactors," March 17, 1992. l SECY-92-120, "NRC Staff Review Schedules for the Westinghouse AP600 and the j General Electric (GE) Simplified Boiling-Water Reactor (SBWR) Designs," April 7, 1992. l SECY-92-127, " Revised Accident Source Tems for Light-Water Nuclear Power i Plants," April 10, 1992. 1 l SECY-92-133, " Draft Safety Evaluation Report for Volume I and Volume III of l the Electric Power Research Institute's Advanced Light-Water Reactor Require-ments Document," April 14, 1992. SECY-92-134, "NRC Construction Inspection Program for Evolutionary and Advanced Reactors Under 10 CFR Part 52," April 15,1992. SECY-92-170, "Rulemaking Procedures for Design Certification," Nay 8, 1992. SECY-92-196, " Development of Design Acceptance Criteria (DAC) for the Advanced Bolling-Water Reactor (ABWR)," May 28, 1992.
" Issues Pertaining to Evolutionary and Passive Light-Water Reactors and Their Relationship to Current Regulatory Requirements," draft Commission paper forwarded to the Commission on February 20, 1992, and made available to the public on February 27, 1992. .
SECY-92-211. "NRC Confirmatory Integral System Testing for the General Electric $8WR Design," June 5, 1992. SECY-92-214. " Development of Inspections, Test, Analyses, and Acceptance Criteria (ITAAC) for Design Certifications," June 11, 1992. SECY-92-219, "NRC-Sponsored Confirmatory Testing of the Westinghouse AP600 Design," June 16, 1992.
" Design Certification and Licensing Policy Issues Pertaining to Passive and Evolutionary Advanced Light-Water Reactor Designs," draft Commission paper forwarded to the Commission on June 25, 1992, and made available to the public on July 1, 1992.
4 - l SECY-92-277, " Final Safety Evaluation Report for Volume II of the Electric Power Research Institute's Advanced Light Water Reactor Requirements Docu- - ment," August 10, 1992. SECY-92-287, " Form and Content for a Design Certification Rule," August 18, 1992. SECY-92-292, " Advance Notice of Proposed Rulemaking on Severe Accident Plant Performance Criteria for Future LWRs," August 21, 1992. SECY-92-294, " Acceptance Review of the Westinghouse Electric Corporation's Application for Final Design Approval and Design Certification for the AP600 Design," August 24, 1992. SECY-92-299, "Gevelopment of Design Acceptance Criteria (DAC) for the Advanced Boiling-Water Reactor (A8WR) in the Areas of Instrumentation and Controls (I&C) and Control Room Design," August 27, 1992. SECY-92-327, " Reviews of Inspections, Test, Analyses, and Acceptance Criteria (ITAAC) for the General Electric (GE) Advanced Boiling Water Reactor (ABWR)," Sept & r 22, 1992. SECY-92-339. " Evaluation of the General Electric Company's (GE's) Test Program to Support Design Certification for the Simplified Boiling-Water Reactor," October 6,1992. SECY-92-368, " Final Rule Amending 10 CFR Part 52," October 29, 1992. SECY-92-381, "Rulemaking Procedures for Design Certification," November 10, 1992. SECY-92-403, " Acceptance Review of GE Nuclear Energy's (GE's) Application for Final Design Approval (FDA) and Design Certification (DC) of the Simplified Boiling-Water Reactor Design (SBWR)," December 3,1992. c h
'i 7 l
i - ' -
- APS3-t ,
VVR ' July 21, 1993 MEMORANDUM FOR: James M. Taylor, Executive Director for Operations FROM: Samuel J. Chilk, Secretary /s/
SUBJECT:
SECY-93-087 - POLICY, TECENICAL, AND LICENSING ISSUES PERTAINING TO EVOLUTIONARY AND ADVANCED LIGET-WATER REACTOR (ALWR) DESIGNS This is to advise you that the Commission (with all Commissioners agreeing) has approved the items for which the staff requested a decision as follows:
- 1. I.E. Fire Protection The Commission approves the staff's position that the passive plants should also be reviewed against the enhanced fire protection criteria approved in the Commission's SRM of i June 26, 1990.
The Commission would like to be kept informed of the staff's ; resolution of the issue related to common-mode failures ! through common ventilation systems.
- 2. I.F. Intersystem Loss-of-Coolant Accident:
The Commission approves the staff's position that the passive plants should also be reviewed for compliance with the intersystem LOCA criteria approved in the Commission's SRM of June 26, 1990. The staff should clarify the intent of the phrase "could not practically be designed to meet such a criterion."
- 3. I.G. Evdrocen control The Commission approves the staff's position that the passive plants should be designed, as a minimum, to the same requirements applied to evolutionary designs. Specifically, i passive plants must:
{ SECY NOTE: This SAN and the vote sheets of all Commissioners will be made publicly availab3e in 10 working days from the date of this SRM. (SEOY-93-087 was i previously released to the, publia) .
1 o accommodate hydrogen generation equivalent to a 100% metal-water reaction of the fuel cladding; o limit containment hydrogen concentration to no greater than 10%; and o provide containment-wide hydrogen control (such as ignitors or inerting) for severe accidents. The Commission approves the staff's clarification, as expressed at the Commission bristing, that the possible use of passive autocatalytic hydrogen recombiners should not be precluded from consideration a priori. The staff is cautioned to consider carefully the relatively slow time response of autocatalytic recombiners as a possible impediment to their efficiency. The staff's resolution of the issue of the placement of hydrogen control devices should be provided to the ACRS and the Commission for information.
- 4. I.E. Core Debris coolabilitvr The Commission approves the staff's position that both the evolutionary and passive ALWR designs meet the following criteria:
o Provide reactor cavity floor space to enhance debris spreading.
)
o Provide a means to flood the reactor cavity to assist in the cooling process. o Protect the containment liner and other structural members 4ith concrete, if necessary. o Ensure that the best estimate environmental conditions (pressure and temperature) resulting from core-concrete
' interactions do not exceed Service Level C for steel containments or Reactor Load Category for concrete containments, for approximately 24 hours., Ensure that the containment capability has margin to accommodate uncertainties in the environmental conditions from core-concrete interactions.
With regard to the 0.02m 2 /MWt reactor vessel cavity floor area, the staff should continue its research activities and supporting analyses, as documented in its May 19, 1993 l
~
letter to the ACRS. I I 2 ,
i , i j With respect to the containment response to ex-reactor vessel core debris, the staff should not limit licensees to ! only one method for addressing containment responses to i severe accident events but also permit other technically
- justified means for demonstrating adequate containment j response.
l 5. I.I. Rich Pressure core Melt Eieotions i i The Commission approves the staff's position for the general criteria that the evolutionary and passive LWR designs: } l J o provide a reliable depressurization system; and I o provide cavity design features to decrease the amount of ejected core debris that reaches the upper containment.
- 6. I.J. containment Performancer The recommendations on containment performance, as outlined in SECY 93-087, could be read to imply that the staff is no longer proposing to use the concept of conditional ,
I containment failure probabilities (CCFP). However, based on discussions held during the commission meeting on this subject, the staff informed the Commission that it intends to continue to apply the 0.1 CCFP in implementing'the Commission's defense in depth regulatory philosophy and the commission's policy on Safety Goals. , j Therefore, the Commission approves the staff's position to use the following deterministic containment performance goal ; in the evaluation of the passive ALWRs as a complement to ' the CCFP approach approved by the Commission in its SRM of l June 26, 1990: 1 "The containment should maintain its role as a reliable, leak-tight barrier (for example, by ensuring that containments stresses do not exceed ASME Service Level C limits for metal containments, or Factored Load Category for concrete containments) for approximately 24 hours following the onset of core damage under the more likely severe accident challenges and, following this period, the containment should continue to provide a barrier against the uncontrolled release of fission products." The Commission approves the staff's interim approach subject to the staff's review and recommendations resulting from public comments on the " Advance Notice of Proposed Rulemaking on Severe Accident Plant Performance Criteria for Future ALWRs."
' 3 ,
\
- I j
i j
- 7. I.E. Dedicated contain= ant vant Penetratient l The Commission approves the staff's position that the need i
for a containment vent for the passive plant designs should be evaluated on a design-specific basis. l 8. I.L. Bruinment Survivabilityt The Commission approves the staff's position that the e ' passive plant design features provided only for severe-accident mitigation need not be subject to the environmental qualification requirements of 10 CFR Section 50.49; quality assurance requirements of 10 CFR Part 50, Appendix B; and redundancy / diversity requirements of 10 CFR Part 50, Appendix A.
- 9. I.M. Elimination of Operatina-Basis Earthauake ROBE):
The Commission approves the staff's recommendation to account for earthquake cycles in the fatigue analyses of piping systems performed until the new guidance is issued, using two SSE events with 10 maximum stress cycl 6s per event (20 full cycles of the maximum SSE stress range). Alternatively, the number of fractional vibratory cycles equivalent to that of 20 full SSE vibratory cycles may be used (but with an amplitude not less than one-third of the l maximum SSE amplitude) when derived in accordance with l Appendix D of IEEE Standard 344-1987. ' The Commission approves the staff's recommendation that the effects of anchor displacements in the piping caused by an SSE be considered with the Service Level D limit. The Commission approves the staff's recommendation to eliminate the'OBE from the design of systems, structures, and components. When the OBE is eliminated from the design, no replacement earthquake loading should be used to establish the postulated pipe rupture and leakage crack locationn. The Commission approves the staff's recommendation that the mechanistic pipe break and high-energy leakage crack locations determined by the piping high stress (without the OB'E) and fatigue locations may be used for equipment environmental qualification and compartment pressurization purposes. The commission agrees that with the elimination of the CBE, two alternatives exist that will essentially maintain the requirements provided in IEEE Standard 344-1987 to qualify equipment with the equivalent of five OBE events follcued by , one SSE event (with 10 maximum stress cycles per event). Of
- 4 . 1
. .. . . . - . .. .1. .
-~.
( -. _- _ i. i j these alternatives, the equipment should be qualified with i five one-half SSE events followed by one full SSE event. j Alternatively, a number of fractional peak cycles equivalent to the maximum peak cycles for five one-half SSE events may j be used in accordance with Appendix D of IEEE Standard 344-j 1987 when followed by one full SSE. 1 The Commiscion agrees that the above requirements should ! also apply to passive ALWRs. 4 l The Commission understands that the OBE will continue to be used as a threshold criterion for conducting inspections ! following an earthquake event. The staff should keep the 4 Commission and the ACRS informed as the staff's further
- analysis and review proceed.
- 10. I.M. Inservice Testina of Pumas and Valvest I The Commission has no objection to the staff's position, but i
understands that further elaboration on this issue will be l forthcoming from the staff. l 11. II.A. Industry codes and standards: i ) The Commission approves the staff's position that consistent
- with past practice, that staff will review both evolutionary i and passive plant design applications using the newest codes j and standards that have been endorsed by the NRC.
i Unapproved revisions to codes and standards will be reviewed. on a case-by-case basis.
- 12. II.D. Leak Before Break The Commission approves the staff's recommendation that the
- leak before break approach should be applied to both the evolutionary and the passive ALWRs seeking design certification under 10 CFR Part 52. This approval should bo
{' limited to instances in which appropriate bounding limits are established using preliminary analysis results during the design certification phase and verified during the COL phase by performing the appropriate ITAAC. ,
- 13. II.E. Classification of Main Steamlines in Boilina Water Reactors:
The commission approves the staff's position that neither the main steam drain and bypass lines from the first valve up to the condenser inlet, nor the piping between the turbine stop. valve and the turbine inlet.should be classified as safety-related or as seismic Category I. Rather, these lines should be analyzed using a dynamic seismic analysis to demonstrate structural integrity under 5 ,
4 i t h SSE loading conditions. The turbine stop, control, and i bypass valves and the main steam lines from the turbine i control valves to the turbine shall meet all of the quality j group and quality assurance guidelines specified in SRP Section 3.2.2, Appendix A. Further, that seismic analyses ! be performed to ensure that the condenser anchoragas and the i piping inlet nozzle to the condenser are capable of j maintaining their structural integrity during and after the l SSE. l 4 The Commission approves the above-described approach to resolve the main steamline classification for both 2 evolutionary and passive ALWRs. 4 14. II.F. Tornado Design Basist i
- The Commission approves the staff's position that a maximum
] tornado wind speed of 482 km/hr (300 mph) be used in the i design-basis tornado employed in the design of evolutionary 3 and passive ALWRs. I
- 15. II.H. Containment Leak Rate Testinas
, The Commission approves the staff's position that until the i rule change proceedings for Appendix J of 10 CFR Part 50 are l completed, the maximum interval between Type C leakage rate tests for both evolutionary and passive plant designs should be 30 months, rather than the 24 months maximum interval currently required in Appendix J to 10 CFR Part 50.
- 10. II.I. Post-Accident Samplina System (PASS): I The Commission approves the staff's position as modified below.
The Commission approves the staff's recommendation that the post-accident sampling systems for evolutionary and passive ALWRs of the pressurized water reactor type be required to have the capability to On:ly:: determine the aross amount of ; dissolved gases (not necessarily a pressurized sample) and ) ehnee4de in ::: rd:::: vith th: r:7.:ir:::nt: as an acceptable means of satisfying the intent of 10 CFR ; 50.34 (f) (2) (viii) and Item II.B.3 of NUREG-0737. The commission agrees that the time for taking these samples can be extended to 24 hours following the accident. The Commission agrees that for evolutionary and passive ALWRs of the boiling water reactor type,.there would be no i need for the post-accident sampling system to analyze l dissolved gases.
. 6 ,
... w. .. - ---
i
^-
a - - .7..---.-----.------------- j . 1 I The Commission approves the deviation from the requirements 1 of Item II.B.3. of NUREG-0737 with regard to requirements i for sampling reactor coolant for boron concentration and ! radioactivity measurements using the post-accident sampling i system in evolutionary and passive ALWRs. The modified l requirement would require the capability to take boron [ concentration samples and radioactivity measurements 8 hours j and 24 hours, respectively, following the accident. ! The Commission approval is based on the fact that the PASS l system is an existing requirement and on the belief that a relatively simple system can be designed to meet the modified requirement. It is the Commission's understanding 4 that a system can be designed which is simple, does not require chemical analysis of the gases in solution, and will ! provide the reactor operator information as to whether l significant amounts of non-condensible gases exist in the , reactor coolant. t i 17. II.M. site-specific Probabilistic Risk Assessments and j n==1vsis of arternal Events l 1 j The Commission approves, in part, and disapproves, in part, the staff's position on site-specific probabilistic risk assessment and analysis of external events, as listed below. l j The Commission approves the position that the analyses submitted in accordance with 10 CFR 52.47 should include an assessment of internal and external events. The Commission disapproves the staff's recommendation to use two. times the Design Basis SSE for margins-type assessment of seismic events. The Commission approves the use of 1.67 times the Design Baris SSE for a margin-type assessment of seismic events. The Commission approves the following staff recommendation, as modified: PRA insights will be used to support a margins-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 containment failures up to approximately one and two-thirds the around motion acceleration of the Desien Basis SSE. The Commission approves the staff's position that the simplified probabilistic methods, such as but not limited to EPRI's FIVE methodology, will be used to evaluate fires. i 7
~~
The Commission approves the staff's position that traditional probabilistic techniques should be used to evaluate internal floods. The Commission approves the staff's position that the ALWR vendors should perform bounding analyses of site-specific external events likely to be a challenge to the plant (such as river flooding,rstorm surge, tsunami, volcaniam, high winds, and hurricanes). The commission approves the staff's position that when a site is chosen, its characteristics should be compared to those assumed in the bounding analyses to ensure that the site is enveloped. The commission approves the staff's position that if the site is enveloped, the COL applicant need not perform further PRA evaluations for these external events. The COL applicant should perform site-specific PRA evaluations to address any site-specific hazards for which a bounding analysis was not performed or which are not enveloped by the bounding analyses to ensure that no vulnerabilities due to siting exist.
- 18. II.O. Defense naminst co==en-Mode Failures in Dicital Instrumentation and control systa==:
The Commission approves, in part, and disapproves, in part, the staff's recommendation. The Commission has approved a revised position, as follows:
- 1. The applicant shall assess the defense-in-depth and diversity of the proposed instrumentation and control system to demonstrate that vulnerabilities to common-mode failures have adequately been addressed.
- 2. In performing the assessment, the vendor or applicant shall analyze each postulated common-mode failure for each event that is evaluated in the accident analysis section of the safety analysis report (SAR) usina best-Antimate methods. The vendor or applicant shall demonstrate adequate diversity within the design for each of these events. 1
- 3. If a postulated common-mode failure could disable a l
safety function, then a diverse means, with a documented basis that the diverse means is unlikely to be subject to the same common-mode failure, shall be required to perform either the same function or a different function. The diverse or different function may be nerformed by a non-safety system if the system 8 ' l l t
. . . . _ . - . . . - - -- ~*
~
is of sufficient cuality to cerform the necessary function under the associated event conditions.
- 4. A set of c fety grade displays and controls located in the main control room shall be provided for manual, system-level actuation of critical safety functions and monitoring of parameters that support the safety functions. The displays and controls shall be independent and diverse from the safety computer system l
identified in items 1 and 3 above. The staff's position has been modified in essentially two respects: First, inasmuch as common mode failures are beyond design-basis events, the analysis of such events should be on a best-estimate basis. Second, the staff indicates in its discussion of the I third part of its position that "The diverse or different function may be performed by a non-safety system if the system is of sufficient quality to perform the necessary function under the associated event conditions." Therefore, this clarification has been added to the fourth part of the staff's position (which refers to a subset of the safety functions referred to in the third part) by removing the safety grade requirement. Further, the remainder of the discussion under the fourth part of the staff position is highly prescriptive and detailed (e.g., "shall be evaluated," "shall be sufficient," shall be hardwired," etc.). The Commission approves only that such prescriptiveness be considered as general guidance, the practicality of which should be determined on a case-by-case' basis.
- 19. II.R. Steam Generator Tube Ruoturest II.R.1. Multiple Steam Generator TUAe Ruotures:
The Commission approves the staff's position to require that analysis of multiple steam generator tube ruptures (STGRs) involving two to five steam generator tubes be included in the application for design certification for the passive PWRs. The Commission understands that, as discussed in the Commission meeting on this SECY paper, since the steam generator multi-tube rupture event is beyond the design basis requirements for PWRs, realistic or best-estimate analytical assumptions may be used to assess plant responses.
' 9 .
c - l ) II.R.2. Containment Rvnans Potential Resultina From SGTRs i i i The Commission approves the staff's recommendation that the applicant for design certification for a passive or j evolutionary PWR assess design features to mitigate the amount of containment bypass leakage that could result from steam generator tube ruptures.
- 20. II.T. control Room ammunoiator (11 arm) Reliability l . The Commission approves the staff's recommendation that the j alarm system for AIMRs should meet the applicable EPRI 1 requirements for redundancy, independence, and separation.
i In addition, alarms that are provided for manually I controlled actions for which no automatic control is { provided and that are required for the safety systems to
- accomplish their safety functions, sha.1.1 meet the applicable
- requirements for Class 1E equipment and circuits.
s
- 21. III.R. Role of the Passive Plant control Room Operator
! The Commission approves the staff's recommendation that i sufficient man-in-the-loop testing and evaluation must be i performed. In addition, a fully functional integrated control room prototype is likely to be necessary for passive plant control room designs to demonstrate that functions and
- tasks are properly integrated into the man / machine
- interface.
Finally, the staff and industry should meet to ensure a common i understanding of the requirements such that industry's design , activities are appropriately directed to comply with the j requirements. The Commission commends the staff for a job well done on the j. highly complex technical issues presented in this paper. ! cc: The Chairman I Commissioner Rogers Commissioner Remick l Commissioner de Planque , 3 ocC IG l ACRS ! OCA OPA 1 10 , l
PpR
/ o UNITED STATES
!' o NUCLEAR REGULATORY COMMISSION { ,I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
@ W ASHINGTON, D. C. 20666 April 22, 1996 The Honorable Shirley Ann Jackson Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PROPOSED REVISIONS TO 10 CFR PARTS 50 AND 100 AND PROPOSED REGULATORY GUIDES RELATING TO REACTOR SITE . CRITERIA 9 During the 430th meeting of the Advisory Committee on Reactor Safeguards, April 11-13, 1996, we reviewed the proposed revisions ' to reactor siting regulations and associated Regulatory Guides and Standard Review Plan sections. Our Subcommittee on Extreme External Phenomena reviewed this matter during a meeting on April l 3, 1996. During this review, we had the benefit of discussions i with representatives of the NRC staff, Westinghouse Electric Corporation, and the Nuclear Energy Institute. We also had the benefit of the document referenced. The staff has proposed final revisions to 10 CFR Parts 50 and 100 and a new Appendix S to Part 50 that deal with both seismic and source term issues for future plants and sites. Many of the implementation details will be found in new Regulatory Guides and < in Standard Review Plan sections. The existing requirements of 10 l CFR Part 100 and its Appendix A will remain in effect for operating l plants. l 1 We recommend that the proposed final rule dealing with the seismic aspects be issued. 1 The proposed final rule requires that any individual, located at any point on the exclusion area boundary for any two-hour period following the postulated release of the fission products, not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE). Similarly, an individual located at the outer boundary of the low population zone (LPZ), who is exposed to the radioactive cloud resulting from the release of the postulated fission products (during the entire period of its passage), not receive a dose in excess of 25 ren TEDE. Consistency within the body of NRC regulations is most desirable. We recommend that careful definitions of the TEDE limits that are mindful of organ dose weighting factors found in 10 CFR Part 20 be included in the final rule. b b h .
__y.._.._ _ _ __.._._ _ _ _ _-_ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ - A /N l [ Radiological doses are to be evaluated over a two-hour period. The proposed final rule states that the evaluation should be over the two-hour period of maximum dose. The Office of Nuclear Regulatory Research (RES) has a differing view and recommends that the proposed final rule be modified from any two-hour period after release of fission products (referred to as the " worst" two hours) to a period of two hours commencing with fuel failure (referred to as the "first" two hours). RES believes that the use of the worst two-hour period in the dose calculation is not justified by risk considerations and could lead to increased costs for future licensees with no commensurate gain in safety.
)
The staff supporting the proposed rule states that (1) the proposed j licensing framework would provide a relaxation of engineered safety I feature (ESF) performance requirements commensurate with updated source term and radiological insights, (2) the regulatory j requirements for determination of in-containment radioactive , material during the two-hour dose evaluation period would be ; consistent and capable of handling designs substantially different from those analyzed in NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants," (3) the analysis would be easy to perform and reproducible with confidence, and (4) the technical , l bases and analytical methods would be defensible. While the j i revised dose evaluation in 10 CFR 50.34 is intended for future l plants, the staff is concerned that a current licensee might seek ; i to use it to remove or disable existing fission product cleanup i systems. .This could markedly change the risk profils of the plant ! from that which was licensed. ' We are not persuaded by the rationale provided by RES in favor of l the first two-hour dose calculation. We agree with the position l taken in the proposed final rule, and recommend that the rule and ! the associated Regulatory Guides and SRP sections be issued. l Sincerely, J 5. /W T. S. Kress Chairman
REFERENCE:
Memorandum dated March 6,1996, from T. P. Speis, Office of Nuclear Regulatory Research, NRC, to J. T. Larkins, ACRS, transmitting Revisions to 10 CFR Part 100, Reactor Site criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections i
A>93-2
? > R.
p* "%
!Y n .j
\...../
RULEMAKING ISSUE May 24, 1996 (Affirmation) secy_96-118 IQ: The Commissioners f!gi: James M. Taylor Executive Director for Operations
SUBJECT:
AMENDMENTS TO 10 CFR PARTS 50, 52, ANG 100, AND ISSUANCE OF A NEW APPENDIX S TO PART 50 PURPOSE: To obtain tommission approval to publish a final rule to amend reactor siting requirements in 10 CFR Parts 50, 52, and 100, including the establishment of a new Appendix S to 10 CFR Part 50, for use by future applicants.
SUMMARY
This paper and accompanying attachments present, for Commission approval, a final rule to amend 10 CFR Parts 50, 52, and 100, and establish a new Appendix S to 10 CFR Part 50. These amendments to the regulations revise basic reactor site criteria and reflect advancements in the earth sciences and earthquake engineering. Two changes to Part 100 are included in this rule. The title of Subpart A is added to include the effective date of this final rule; this action will preserve the licensing basis for existing plants. Subpart A and Appendix A to Part 100 are identical to the present rule. Subpart B, applicable to future plants, is added to Part 100 and contains basic nonseismic site criteria, without numerical values, in a new s 100.21, "Nonseismic Siting Criteria." ! Seismic criteria are included in a new s 100.23, " Geologic and Seismic Siting Factors." Revisions to 10 CFR Part 50 contain source term and dose criteria (s 50.34) and earthquake engineering criteria (new Appendix S). The revision to 10 CFR 50.34 reflects the staff recommendation and rationale for the revised dose criteria to be used to judge the applicability of plant designs. l
Contact:
NOTE: TO BE MADE PUBLICLY AVAILABLE AT Leonard Soffer, EDO COMMISSION MEETING ON JUNE 3, 1996 415-1722 Dr. Andrew J. Murphy, RES 415-6010 pf ,n { . n[ U= 1 1 NI [ ph
7 .[ T i The Commissioners 2 BACKGROUND: On April 12, 1962, the Atomic Energy Commission (AEC) issued 10 CFR Part 100,
" Reactor Site Criteria" (27 8 3509). On November 13, 1973, the AEC issued i Appendix A to 10 CFR Part 100, " Seismic and Geologic Siting Criteria for ;
Nuclear Power Plants," (38 8 31279). j A proposed rule to revise Part 100, Appendix A to Part 100, and sections of Part 50 was published for comment on October 20, 1992 (57 8 47802). The l proposed rule change combined two separate initiatives dealing with non-seismic and seismic issues, and included a minimum distance to the exclusion 1 area boundary of 0.4 miles, guideline limits for population density, and I required both probabilistic and deterministic seismic hazard evaluations. The i comment period, extended twice, expired on June 1, 1993. Extensive comments, ' both domes, tic and international, were received. l The Consission was briefed on August 3,1993, on the status of the proposed , rule and the nature of the comments received. In an SRM dated August 12, l 1993, the Commission raised several concerns regarding the prescriptive aspects of the proposed revisions to Part 100 as well as its form and content. In response, the staff prepared an options paper, SECY-94-017, dated January 26, 1994. In an SRM dated March 28, 1994, the Commission approved the staff recommendations. However, due to the substantive nature of the changes to be made to the rule the Commission stated that both parts were to be resubmitted 1 for Commission review and reissued for public comment before developing the final rulemaking. Outlines of the draft regulatory guides and standard review plan section were to be submitted to the Commission for review, to demonstrate how the basic site criteria are to be implemented. The draft regulatory guides and standard review plan section were to be issued for public comment after receiving Commiss; ion approval of the outlines. The second proposed revision to these regulations was published for public comment on October 17, 1994 (59 FR 52255). On February 8, 1995, the NRC extended the comment period to allow interested persons adequate time to ; provide comments on staff guidance documents (60 FR 7462). On ' February 28, 1995, a notice of availability was published for the five draft regulatory guides and three draft standard review plan sections that were , developed to provide guidance on meeting the proposed regulations j (60 FR 10880). The comment period for the proposed rule was extended to > May 12, 1995 (60 FR 10810). Included in this package are the Federal Register notice for the final rule (Attachment 1), the resolution of public comments on the proposed seismic and earthquake engineering criteria for nuclear power plants (Attachment 2), the ACRS letter on the rulemaking (Attachment 3), a draft public announcement (Attachment 4), the draft congressional letters (Attachment 5), draft letters to the Speaker of the House of Representatives, President of the Senate, and ) the General Accounting Office (Attachment 6), regulatory analysis ' (Attachment 7), environmental assessment, (Attachment 8), regulatory guidance for general site suitability criteria (Attachment 9), and regulatory guidance and public comment resolution for the seismic and earthquake engineering criteria (Attachments 10-17). L
y s b The Commissioners 3 l DISCUSSION: NON-SEISMIC ASPECTS: Proposed rule The proposed rule issued for comment on October 17, 1994 (FR 59 52255) would retain the use of source term and dose calculations (relocating these to Part
- 50) to determine the distance to the exclusion area boundary (EAB) and the 4 size of the outer radius of the low population zone (LPZ). The proposed dose i criteria would require that an individual located at any point on the boundary of the exclusion area for any two-hour period following the onset of the postulated fission product release not receive a dose in excess of 25 rem total effective dose equivalent (TEDE). Similarly, an individual located at the outer boundary of the LPZ for the entire period of the cloud passage (taken to be 30 days) must not receive a dose in excess of 25 rem TEDE.
1 Section 100.21 proposed to contain basic site criteria without any numerical I values. With regard to population density, the proposed rule stated that: Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining 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, economic, or other factors, which may result in the site being found acceptable. Revision 2 of Regulatory Guide 4.7 (draft Regulatory Guide DG-4004) would i contain guidance on preferred population density as follows: l l A reactor preferably should be located such that at the time of initial I site approval and within about 5 years thereafter, the population density, including weighted transient population, averaged over any radiai distance out to 20 miles (cumulative population at a distance divided by the circular area at that distance) does not exceed 500 I persons per square mile. A reactor should not be located at a site whose population density is well in excess of the above value. If the population density of the proposed site exceeds, but is not well l in excess of the above preferred value, an analysis of alternative sites should be conducted for the region of interest with particular attention to alternative sites having lower population density. However, l consideration will be given to other factors, such as safety, environmental, or economic considerations, which may result in the site with the higher population density being found acceptable. Examples of such factors include, but are not limited to, the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail or highway access, shorter transmission line requirements, or less environmental impact upon undeveloped areas, wetlands, or endangered species. l
a s > \ l j The Commissioners 4 Public Comments: I Eight organizations or individuals commented on the nonseismic aspects of the second proposed revision. A summary of the public comments received was transmitted to the Connission in a memorandum dated June 19, 1995. The first proposed revision issued for comment in October 1992 elicited strong comments 1 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 would delete these from the rule by providing guidance on population density in a Regulatory Guide and determining the distance to the EAB and LPZ by use of source term and dose calculations. The rule would contain basic site criteria, without any numerical values. Several connenters 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 commented that the NRC had retreated from decoupling siting and design in response to the comments of foreign entities. Most comments 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 period, and the question of whether an organ capping dose should be adopted. Virtually all commenters 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 commenters 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 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 commenters opposed the use of an organ capping dose. The only public interest group that commented did not object to the use of TEDE, favored the proposed dose value of 25 rem, and supported an organ capping dose. Final Rule: 10 CFR 50.34 The final rule makes no changes that were not presented in the proposed rule. The final rule would require, as in the proposed rule, that an individual located at any point on the boundary of the exclusion area for any two hour period following onset of the postulated fission product release, not receive a radiation dose in excess of 25 rem total effective dose equivalent (TEDE).
- Similarly, an individual located at the outer boundary of the low population zone (LPZ), who is exposed to the rsdioactive cloud resulting from the postulated fission product release (during the entire period of its passage) not receive a dose in excess of 25 rem TEDE.
, y
= b The Commissioners 5 The staff recommends adoption of a dose acceptance criterion of 25 rem TEDE based upon consideration of the risk of latent cancer fatality, as noted in
-the Statement of Considerations that accompanied the proposed rule. The staff also notes that, in terms of occupational dose, Part 20 permits a once-in-a-lifetime planned special dose of 25 rem TEDE, and that this value provides a ,
useful perspective with regard to doses that ought not to be exceeded for radiation workers. In addition, EPA guidance sets a limit of 25 rem TEDE for workers performing emergency service such as lifesaving or protection of large populations. Because the TEDE concept accounts for the contribution from all body organs, the staff recommends that no additional organ " capping" dose be required. A number of comments were received indicating that the proposed value of 25 rem TEDE represented a more restrictive criterion than the current values of
- 25 rem to ,the whole body and 300 rem to the thyroid. These commenters noted that use of the organ weighting factors of 10 CFR Part 20 of I and 0.03 for the whole body and the thyroid gland, respectively, would yield a TEDE dose of 34 rem. This is because the organ weighting factors of Part 20 include other effects (e.g., genetic) in addition to latent cancer fatality. The argument that a dose criterion of 25 rem TEDE represents a tightening of the current dose criteria, 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 licensing reviews, and that all operating plants would te able to meet a dose criterion of 25 rem TEDE. ,
Hence, the staff concludes that use of the organ weighting factors of Part 20 together with a dose ' criterion of 25 rem TEDE, in practice, represents a relaxation rather than a tightening of the dose criterion. With respect to the two hour evaluation period, the staff continues to support the regulatory approach for the two hour dose evaluation period that was articulated in the proposed revision published on October 17, 1994 (any two hour period). The Office of Nuclear Regulatory Research has a differing view , and recommends a dose evaluation period consisting of the first-two hours i following the onset of core damage. A discussion of the issues involved regarding the two hour dose evaluation period, i.e., any two hour period vs. first two hour period, was provided to the Commission in a memorandtm to Chairman Jackson from James M. Taylor dated April 30, 1996. 10 CFR 100.21 No comments were received that proposed changes to the recu14 tion and no changes are recommended by the staff in the final rule. Revision 2 of Reaulatory Guide 4.7 (draft Reaulatory Guide DG-4004) One comment, while supporting the concept of environmental justice, expressed concern regarding subjective phrases and potential implementation and recommended that the environmental justice provision be deleted from this version of the Guide until more detailed guidance becomes available. The staff recognizes that detailed implementation guidance may not yet be available in this area, but recommends that the environmental justice provision be retained in issuing this Guide in final form. ,
_ _ _ _ _ _ ~
, b ;
The Commissioners 6 Reaulatory Guides 1.3 and 1.4 These Regulatory Guides describe the methodology currently used in performing the dose calculations. The staff plans to develop updated Guides to be consistent with the final rule, once the final rule is approved. I SEISMIC ASPECTS: Proposed Rule: Because no significant changes were made to the regulations published for public comment this discussion will focus on the differences between the current (Appendix A to Part 100) and final regulations (s 100.23 and Appendix S to Part 50) and staff resolution of the public comments. Final Rule': Because the criteria presented in the regulation will not be applied to existing plants, the licensing bases for existing nuclear power plants must remain part of the regulations. Therefore, the criteria on seismic and geologic siting are designated as a new s 100.23 and added to the existing body of regulations in 10 CFR Part 100. In addition, earthquake engineering criteria are located in 10 CFR Part 50, _in a new Appendix S. Because Appendix S is not self executing, applicable sections of Part 50 (s50.8 and
$50.34) are revised to reference Appendix S. Conforming amendments to 10 CFR Parts 52 and 100 are also made. Sections 52.17(a)(1), 52.17(a)(1)(vi), 100.8, ;
and 100.20(c)(1) and (3) are amended to note s 100.23 or Appendix S to Part )
- 50. ;
i Geoloaic and Seismic Sitina j The regulations and guidance documents reflect new information and'research results, as well as comments from the public. In response to the August 12, 1993, SRM pertaining to the prescriptive aspects of the first proposed revisions to Part 100 as well as its form and content, the final regulation only contains the basic requirements. The detailed guidance similar to that contained in Appendix A to 10 CFR Part 100 has been removed to guidance documents. Thus, the new regulation (s 100.23) contains: (a) required definitions, (b) a requirement to determine the geological, seismological, and engineering characteristics of the proposed site, and (c) requirements to determine the Safe Shutdown Earthquake Ground Motion (SSE), to determine the potential for surface deformation, and to determine the design bases for seismically induced floods and water waves. Detailed guidance, that is, procedures acceptable to the NRC staff for meeting the requirements, is contained in Regulatory Guide 1.165, " Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motion," (Draft was DG- .:2) . NRC staff review guidelines is provided in Standard Review Plan (SL , Section 2.5.2, " Vibratory Ground Motion," Revision 3. Two other SRP secti;.3, 2.5.1, " Basic Geologic and Seismic Information," and 2.5.3, " Surface Faulting," are also revised to assure consistency among the rule, SRP Section 2.5.2, and Regulatory Guide 1.165.
<.-n , , - . - - . _ . - . . _ . . - _ _ . - - _ . - - . . - - - - . - - . - i The Commissioners 7 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, 3
relies on a " deterministic" approach. Using this deterministic approach, an , 3 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 1 l can affect the site, locates the postulated earthquake according to prescribed j j rules, and then calculates ground motions at the site. -
4 Although this approach has worked reasonably well for the past two decades, in I the sense that SSEs for plants sited with this approach are judged to be ! suitably conservative, the approach has not explicitly recognized l uncertainties in geosciences parameters. Because of the uncertainty about
- earthquake phenomena (especially in the eastern United States), there have 4 often been differences of opinion and differing interpretations among experts 4 as to the largest earthquakes to be considered and ground-motion models to be .
j used, thus'often making the licensing process relatively cumbersome. ! l Over the past decade, analysis methods for incorporating these different I interpretations have been developed and used. These "probabilistic" methods J have been designed v 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 to not only incorporate
- different models 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 j providing an explicit expression for the uncertainty in the ground motion
- estimates and a means of assessing sensitivity to various input parameters.-
j Another advantage of the probabilistic method endorsed in Regulatory Guide { 1.165 is the target exceedance probability'is set by examining the design , j ba.ses of more recently licensed nuclear power plants resulting in a more uniform level of safety from sits to site. l l The revision to the regulation now explicitly recognizes that there are , ! inherent uncertainties in establisiing the seismic and geologic design l parameters and allows for the option of using a probabilistic seismic hazard
- methodo!ogy capable of propagating uncertainties as a means to address these j uncertainties. The rule. further recognizes that the nature of uncertainty and
- the appropriate approach to accour.t for it depend greatly on the tectonic j regime and parameters, such 3s. the knowledge of seismic sources, the j existence of historical and recorded data, and the understanding of tectonics.
i Therefore, methods other than the probabilistic methods, such as sensitivity 1 analyses, may be adequate to account for uncertainties for some sites. i The key elements of the approach exemplified in Regulatory Guide 1.165 and l Standard Review Plan Section 2.5.2 are described below in steps (a) through ! (g). It should be noted that by this rulemaking the Commission would be ' endorsing implicitly the expert e11 citation processes, including the method for aggregation of expert opinion, described in (1) NUREG/CR-5250, " Seismic 4 Hazard Characterization of 69 Nuclear Plant Sites East of the Rocky i Mountains," (2) NUREG-1488, " Revised Livermore Seismic Hazard Estimates for i Sixty-Nine Nuclear Power Plant Sites East of the Rocky Mountains," and (3) l Electric Power Research Institute report NP-6395-D, "Probabilistic Seismic i Hazard Evaluations at Nuclear Power Plant Sites in the Central and Eastern I- _- _ _ _
g The Commissioners 8 United States: Resolution of the Charleston Earthquake Issue," which produced the probabilistic seismic hazard assessment methods.
- a. Conduct site-soecific and reaional aeoscience investiaations, j These investigations are performed to determine specific characteristics of the proposed site, such as, the presence or absence of potential seismic sources, capable faults at or near the site, characterization of the rock and soil strata, earthquake history of the site and environs, etc. In addition to characterizing the site, these data are needed to verify that regional characteristics used in the Lawrencs Livermore National Laboratory (LLNL) or the Electric Power Reseate.h Institute (EPRI) i probabilistic seismic hazard assessments (PSHA)-are valid for the proposed site.
- b. Taraet exceedance probability is set by examinine the desian bases of more recentiv licensed nuclear nower plants.
The target exceedance probability is the median annual probability of exceeding the Safe Shutdown Earthquake (SSE) for operating nuclear power plant that were designed to Regulatory Guide 1.60 or to a similar spectrum. This value has been determined to be 1E-5/ year.
- c. Determine if information' from aeoscience investiaations chanae probabilistic results. ;
The applicant conducts an evaluation that demonstrates that the ' data obtained from the site investigations (Step a. above) do not provide information that would necessitate revision of the seismic ; sources used in the existing seismic hazard studies and their : characteristics-or attenuation models. ! 1
- d. Conduct orobabilistic seismic hazard analysis and determine around motion level correspondina to the taraet exceedance orobability.
The applicant conducts a LLNL or EPRI PSHA for the proposed site to obtain a seismic hazard curve, grour.d acceleration or spectral amplitude vs. annual probability of exceedance. The hazard curve median is deaggregated to determine a seismic event described by- l an average earthquake magnitude and distance (distance from earthquake to the nuclear power plant site) which contributes most to the ground motion level corresponding to the target exceedance probability. This magnitude and distance is then used in subsequent steps to determine site-specific spectral shape.
- e. Determine site-specific spectral shape and scale this shape to the around motion level determined above.
The applicant will use the seismic event of magnitude and distance determined in Step d to develop site-specific spectral shapes in accordance with SRP 2.5.2 procedures and additional guidance provided in the regulatory guide. The SRP procedures, in part, are based on use of seismic recorded motions or ground motion models appropriate for the event, region and site under consideration. I
C 5 i The Commissioners 9 l NRC staff review of around motion. f.
- l. -The NRC staff will review the applicant's proposed SSE ground motion to assure that it takes into account all available data
! including insights and information gained from previous licensing experience. {- g. Undate the data base and reassess probabilistic methods at least every ten ! y.tAtt. j To keep the regulatory guidance on the probabilistic methods and
~ their seismic hazard data base current, the NRC would reassess
{ them at least every ten years and update them as appropriate. i The results of the regional and site-specific investigations must be l considered in the application of the probabilistic method. The current j probabilistic methods (the NRC sponsored study conducted by LLNL or the EPRI
- seismic ha'zard study), are regional studies without detailed information on j any specific location. The specific applicant's geosciences investigations
! are used to update the database used by the probabilistic hazard methodology l to assure that all appropriate information is incorporated. It_ is 'also necessary to incorporate local site geological factors such as stratigraphy and to account for site-specific geotechnical properties in establishing the design-basis ground motion. In order to incorporate local site factors and advances in ground motion attenuation models, ground motion estimates are determined using the procedures that are outlined in Standard Review Plan Section 2.5.2. The NRC staff's approach to evaluating an application is described in SRP , Section 2.5.2. This review takes into account the information base developed in licensing more than 100 plants. Although the premise in establishing the target exceedance probability is that the current design levels are adequate, a staff review assures that there is consistency with previous licensing decisions and that the scientific basis for decisions are clearly understood. This review approach will also assist in assessing the fairly complex regional probabilistic modeling which incorporates multiple hypotheses and a multitude of parameters. Furthermore, this process should provide a clear basis for the staff's decisions and facilitate communication with nonexperts. Earthauake Enaineerina Criteria not associated with the selection of the site or establishment of the Safe Shutdown Earthquake Ground Motion (SSE) have been placed into Part 50. This action is consistent with the location of other design requirements in Part 50. The regulation is a new Appendix S, " Earthquake Engineering Criteria for Nuclear Power Plants," to Part 50. In the current regulation, Appendix A to Part 100, the Operating Basis Earthquake Ground Motion (OBE), the vibratory ground motion that will assure safe continued operation, is one-half the SSE. In Appendix S, this requirement has been replaced with two options: (1) applicant selection of an OBE that is either one-third of the SSE or less, or (2) a value greater than one-third of the SSE. With the OBE level set at one-third or less of the SSE,
.- o ,
i The Commissioners 10 only the SSE is used for design; the OBE only serves the function of an inspection and shutdown level. If the OBE is greater than one-third of the : SSE, the current practice of using both the OBE and SSE for design continues; and in addition, the OBE serves the function of an inspection and shutdown ' level. This change responds to one of the major criticisms with the existing , regulations, that the OBE controls the design of some parts of the plant. For new applications the regulation would treat plant shutdown associated with vibratory ground motion exceeding the OBE (or significant plant damage) as a condition in every operating license. Section 50.54 is revised accordingly. , Related plant shutdown and OBE exceedance guidelines for operating plants are ' being developed separately by NRR. Procedures acceptable to the NRC staff for meeting the requirements in the new regulation, will be contained in three regulatory guides, (a) Regulatory Guide 1.12, " Nuclear Power Plant Instrumentation for Earthquakes," Revision 2 (Draft was DG-1033), (b) Regulatory Guide 1.166, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions" (Draft was DG-1034), and (c) Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event" (Draft was DG-1035). Public Comments Seven letters were received addressing either the regulations or both the regulations and the draft guidance documents. An additionha . 've letters were received addressing only the guidance documents, for a total of 12 comment i letters. ! 10 CFR 100.23 No changes were made to the regulation as a result of the public comments. In I general, the commenters were supportive of the regulation, specifically, the
- removal of prescriptive guidance from the regulation and locating it in regulatory guides or standard review plan sections and the removal of the requirement from the first proposed rulemaking (57 FR 47802) that both deterministic and probabilistic evaluations must be conducted to determine site suitability and seismic design requirements for the site.
A suggestion that for existing sites east _ of approximately 105' west longitude (the Rocky Mountain front), a 0.3g standardized design level be r.odified was not adopted. The NRC has determined that the use of a spectral shape anchored to 0.3g peak ground acceleration as a standardized design level would be appropriate for existing sites based on the current state of knowledge. However, as new information 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. The suggestion to change the regulation to enable an applicant for an operating license already holding a construction permit to apply the amended methodology and criteria in Subpart B to Part 100 was not incorporated. The NRC will address this request on a case-by-case basis rather than through a
The Commissioners 11 generic change to the regulations. This situation pertains to a limited number of facilities in various stages of construction. Some of the issues that must be addressed by the applicant and NRC during the operating license review include differences between the design bases derived from the current and amended regulations (Appendix A to Part 100 and s 100.23, respectively), and earthquake engineering criteria such as, OBE design requirements and OBE shutdown requirements. An explicit statement whether or not s 100.23 applies to the Mined Geologic Disposal System (MGDS) and a Monitored Retrievable Storage (MRS) facility was not added to the regulation or Supplemental Information Section of the rule. Presently, NUREG-1451, " Staff Technical Position on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Repository," notes that Appendix A to 10 CFR Part 100 does not apply 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. The 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 a MGDS until it is superseded through a rulemaking, revision of NUREG-1451, or other appropriate mechanism. Annendix S to 10 CFR Part 50 Support for the NRC position pertaining to the elimination of the Operating Basis Earthquake Ground Motion (0BE) response analyses has been documented in various NRC publications such as SECY-79-300, SECY-90-016, SECY-93-087, and NUREG-1061. 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. In addition, similar activities are being done in foreign countries, such as, Germany. However, one commenter expressed concern about the elimination of OBE response analyses of pressure-retaining components designed to the ASME Boiler and Pressure Vessel Section III rules. Positions pertaining to the elimination of the OBE were proposed in SECY-93-087. Commission approval is documented 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. Item V(B)(5), "Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analysis," to the supplemental information to the regulations was slightly modified to address the noted Concerns. The regulation was not changed to incorporate by nference the American Society of Civil Engineers (ASCE) Standard 4, "Seisinic Analysis of Safety-Related Nuclear Structures and Commentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures." In response to the August 12, 1993, SRM pertaining to the prescriptive aspects of the first proposed revisions to Part 100 as well as its form and content, the final regulation contains only the basic requirements;the detailed guidance is provided in regulatory guides and
,. a
~- ~
The Commissioners 12 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.
l The reference to aftershocks in Paragraph IV(b), Surface Deformation was j deleted. Paragraphs VI(a)(1), " Safe Shutdown Earthquake," and VI(b)(3) of
- Appendix A to Part 100 contain the phrase " including aftershocks." In the 4
proposed regulation the " including aftershocks" phrase was only removed from l the Safe Shutdown Earthquake Ground Motion requirements (Paragraph IV(a)(1) of I l Appendix S to Part 50). ) l Guidance Documents 1 Many of the commenters have provided editorial and technical suggestions that clarified the documents. A few commenters provided more substantive comments required a careful assessment of their implications. For example, based upon public con ~ ment, the NRC staff clarified the procedure in SRP Section 2.5.2 used to assess the adequacy of an applicants submittal. Also, Regulatory Guide 1.165 (Draft was DG-1032) now includes a discussion of how uncertainties in the SSE can be addressed through a suitable sensitivity analysis. In general, no technical changes were made to the staff positions described in , the draft guidance documents. 1 It is anticipated that the notice of availability of the related regulatory guidance and standard review plan sections will be published in the Federal Reaister coincident with the effective date of the final regulations. l COORDINATION: Coordination will be initiated with the Office of Information and Regulatory Affairs (OIRA), Office of Management and Budget concerning whether this final . rule is a " major rule" as defined in Section 804 of the Small Business Regulatory Enforcement Fairness- Act of 1996. The staff believes that this ; action does not meet the statutory definition of a " major rule" and Attachment I has been prepared on this basis. If the OIRA determines that this is a
" major rule," Attachment I will be revised and the final rule will be amended to include a Regulatory Flexibility Analysis. The Offices of Nuclear Reactor Regulation and Administration concur on this Commission Paper. The Advisory i Committee on Reactor Safeguards was briefed and has provided their views (Attachment 3). The Committee to Review Generic Requirements was provided this Commission Paper for review and they have no objection to issuing this rule. The Office of the General Counsel has no legal objection.
RECOMMENDATIONS: That the Commission:
- 1. Anorove publication of the Revisions to the Regulatory Requirements for Reactor Siting (Seismic and Nonseismic) and Earthquake Engineering Criteria in 10 CFR Parts 50, 52, and 100 (Attachment 1) as a final rule.
l 1 l l The Commissioners 13 1 1
- 2. Certify that this rule will not have a significant economic effect on a substantial number of small entities pursuant-to the Regulatory Flexibility Act (5 U.S.C. 605(b)). )
- 3. Egit:
)
- a. The final rule will be published in the Federal Reaister and become effective 30 days after publication.
- b. The reporting and recordkeeping requirements contained in this 1 regulation have been approved by the Office of Management and Budget, OMB approval Numbers 3150-0093 and 3150-0011.
- c. A public announcement (Attachment 4) will be issued when the l
. notice of rulemaking is sent to the Office of the Federal Register.
- d. The appropriate Congressional committees will be informed
, (Attachment 5).
- e. The letters necessary to inform the Speaker of the House of l Representatives, the President of the Senate, and the General 3 Accounting Office of this final rule (as required by the Small
. Business Regulatory Enforcement Fairness Act of 1996) will be transmitted after the rule has been signed by the Secretary of the Commission (Attachment 6).
- f. Copies of the Federal Reaister notice will be distributed to all power reactor licensees. The notices will be sent to other interested parties upon request.
- g. The Chief Counsel for Advocacy of the Small Business Administration will be notified of the Commission's determination, pursuant to the Regulatory Flexibility Act (5 U.S.C. 605 (b)),
that this rule will not have a significant economic effect on a substantial number of small entities.
- h. He availability of the final regulatory guides and standard review plan sections will be published in the Federal Reaister subsequent to the effective date of the final rule.
- i. A copy of " Resolution of Public Comments on the Proposed Seismic
- and Earthquake Engineering Criteria for Nuclear Power Plants" (Attachment 2), will be placed in the Public Document Room and sent to interested parties upon request.
/ ^ ; s kM.Tay f r Fjecutive Director for Operations
The Commissioners 14 Attachments:
- 1. Federal Register Notice of Rulemaking
- 2. Resolution of Public Comments on the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants
- 3. ACRS Letter
- 4. Draft Public Announcement
- 5. Draft Congressional Letters ;
- 6. Draft Letters to the Speaker of the House of Representatives, the President of the Senate, and the General Accounting Office
- 7. Regulatory Analysis
- 8. Environmental Assessment
- 9. Draft Regulatory G'r:de DG-4004 (General Site Suitability Criteria) '
*10. Regulatory Guidt 1.165 (Seismic Sources, Draft was DG-1032)
*11. Standard Review Plan section 2.5.1, Revision 3
*12. Standard Review Plan Section 2.5.2, Revision 3
*13. Standard Review Plan Section 2.5.3, Revision 3
*14. Regulatory Guide 1.12, Revision 2 (Instrumentation, Draft was DG-1033)
*15. Regulatory Guide 1.166 (Plant Shutdown, Draft was DG-1034)
*16. Regulatory Guide 1.167 (Plant Restart, Draft was DG-1035)
*17. Resolution of Public Comments on Draft Regulatory Guides and Standard Review Plan Sections Pertaining to the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants
- Commissioners, SECY, 0GC only Commissioners' comments or consent should be provided directly to the Office !
of the Secretary by COB Monday, June 17, 1996. Commission Staff Office comments, if any, should be submitted to the Commissioners
- NLT June 10, 1996, with an information copy to the Office of the Secretary. If the paper is of such a nature that it requires additional review and comment, the Commissioners and the Secretariat should be apprised of when comments may be j expected. j i
This paper is tentatively scheduled for afft:mation at an Open Meeting during i the week of June 17, 1996. Please refer to the appropriate Weekly Commission l Schedule, when published, for a specific date and time. l DISTRIBUTION: Commissioners OGC OCAA OIG OPA OCA REGIONAL OFFICES EDO SECY
- .. a. . - - . - . - , . . r-.a -a a- .s w .- .- a ax. ,- . - . s --. . * - ,.s--- x.---a.- - - -
4 4 e ATTACHMENT 1 FEDERAL REGISTER NOTICE OF RULEMAKING -
! i FRN-100.R12 5/22/96- [7590-01-P] . 7 i NUCLEAR REGULATORY COMISSION i 10 CFR Parts 50, 52 and 100 , J j RIN 3150-AD93 3 i Reactor Site Criteria
, Including Seismic and Earthquake Engineering Criteria for Nuclear Power Plants and Denial of Petition from Free Environment, Inc. et. al.
~ AGENCY: Nuclear Regulatory Commission.
ACTION: Final rule and denial of petition from Free Environment, Inc. ) et.al. j SUMARY: The Nuclear Regulatory Commission (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 I 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 i separate changes, namely, the source ters and dose considerations, and the ' seismic and earthquake engineering considerations'of reactor siting. The Commission also is denying the remaining issue in petition (PRM-50-20) filed : by Free Environment, Inc. et. al. ! 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 Commission, Washington, DC 20555-0001, telephone (301) 415-6010, concerning the seismic and earthquake engineering aspects and Mr. Leonard Soffer, Office of the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1722, concerning other siting aspects.
SUPPLENENTARY INFORMATION: I. Background. II. Objectives. III. Genesis. IV. Alternatives. V. Najor 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. Summary of Comments on the Proposed Regulations. , A. Reactor Siting Criteria (Nonseismic). l B. Seismic and Earthquake Engineering Criteria. i X. Small Business Regulatory Enforcement Fairness Act XI. Finding of No Significant Environmental Impact: Availability. i XII. ' Paperwork Reduction Act Statement. I XIII. Regulatory Analysis. XIV. Regulatory Flexibility Certification. XV. Backfit Analysis. I. Background The present regulation regarding reactor site criteria (10 CFR Part 100) was promulgated April-12, 1962 (27 FR 3509). NRC staff guidance on exclusion , area and low population zone sizes as well as popnlation density was issued in Regulatory Guide 4.7, " General Site Suitability Criteria for Nuclear Power Stations," published for comment 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 (PRil-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 Commission 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 recommendations 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 ! Rulemaking (ANPRM) regarding revision of the reactor site criteria, which discussed the recommendations of the Siting Policy Task Force and sought public comments. .The proposed rulemaking was deferred by the Commission in December 1981 to await development of a Safety Goal and im) roved research on accident source terms. On August 4, 1986 (51 FR 23044), t1e 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 PRM-100-2 on the basis that it would
+
2
ga . = e . . l unnecessarily restrict NRC's regulatory siting policies and would not result i
- in a substantial increase in the overall protection of the public health and i- safety. The Commission is addressing the remaining issue in PRM-50-20 as part of this rulemaking action.
i Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power-Plants," to 10 CFR Part 100 was originally issued as a proposed regulation on :
- November 25,1971 (36 FR 22601), published as a final regulation on November ,
i 13,-1973,(38 FR 31279), and became effective on December 13, 1973. There have : i been two amendments to 10 CFR Part 100, Appendix A. The first amendment, ! issued November 27,1973 (38 FR 32575), corrected the final regulation by j adding the. legend under the diagram. The second amendment resulted from a i petition for rulemaking (PRM 100-1) requesting that an opinion be issued that 1 would interpret and clarify Appendix A with respect to the determination of ! the Safe Shutdown Earthquake. A notice of filing of the petition was l published on May 14,-1975 (40 FR 20983). The substance of the petitioner's , j proposal was accepted and published as an immediately effective final ! 4 regulation on January 10, 1977 (42 FR 2052). The"first proposed revision to these regulations was published for public comment 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. First, the NRC staff initiated an extension (58 FR 271) from 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 comment period was extended to June 1, 1993 (58 FR 16377). The second proposed revision to these regulations was published for public comment on October 17, 1994 (59 FR 52255). The NRC stated on February 8,1995, (60 FR 7467) that it intended to extend the comment period to allow interested persons adequate time to provide comments 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 comment period for the proposed rule was extended to May 12, 1995 (60 FR 10810). II. Objectives The objectives of this regulatory action are to --
- 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 permit consideration of new technical understandings; and
- 3. Relocate source term and dose requirements that apply primarily to plant design into 10 CFR Part 50.
3
~,---o ,
l J III. Genesis The regulatory action reflects changes that are intended to (1) benefit 1 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 apply 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 final regulation, 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 sub' sequent proceedings, including license amendments 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 i ' requirements in 10 CFR Part 50. i Because the revised criteria presented in this final regulation does not j apply to existing plants, the licensing bases for existing nuclear power j plants must remain a part of the regulations. Therefore, the non-seismic and q seismic reactor site criteria for current plants is retained as Subpart A and i Appendix A to 10 CFR Part 100, respectively. The revised reactor site i criteria is added as Subpart B in 10 CFR Part 200 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 s100.21 to Subpart B in 10 CFR i Part 100. The criteria on seismic and geologic siting is added as a new s100.23 to Subpart B in 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 S is not self executing, applicable sections of Part 50 (s50.34 and s50.54) are revised to reference Appendix S. The regulation also makes conforming amendments to 10 CFR Part
- 52. Section :i2.17(a)(1) is amended to reflect changes in s 50.34(a)(1) and 10 CFR Part 100.
IV. Alternatives The first alternative considered by the Commission was to continue using i current regulations for site suitability determinations. This is not 1 considered an acceptable alternative. Accident source terms and dose calculations currently primarily influence plant design requirements rather l 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 l 3 safety. Further,'significant advances in understanding severe accident i a behavior, including fission product release and transport, as well as in the l 1 l 4 - l
- - , m. - - r v -.-- +
earth sciences and in earthquake engineering have taken place since the y promulgation of the present regulation and deserve to be reflected in the
- regulations.
l The second alternative considered was replacement of the existing l regulation with an entirely new regulation. This is not an acceptable l alte native because the provisions of the existing regulations form 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 i 10 CFR Part 100 and relocating plant design requirements to 10 CFR Part 50 l j while retaining the existing regulation was chosen as the best alternative. ) i The public will benefit from a clearer, more uniform, and more consistent I ! licensing process that incorporates updated information and is subject to ) i fewer interpretations. The NRC staff will benefit from improved regulatory i implementation (both technical and legal), fewer interpretive debates, and t increased regulatory flexibility. Applicants will derive the same benefits in , ' addition t~o avoiding licensing delays caused by unclear regulatory l requirements. i V. MAJOR CHANGES i. l A. Reactor Siting Criteria (Nonselsnic). Since promulgation of the reactor site criteria in 1962, the Commission has !- approved more than 75 sites for nuclear power reactors and has had an l 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 I operational experience, a great deal of insight has been gained regarding the l 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 i 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 i on radioactive material releases to the environment under severe accident I conditions have all confirmed that present commercial 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 l Safety Study" (WASH-1400), published in 1975, many Probabilistic Risk Assessment (PRA) studies conducted on individual plants as well as several specialized studies, and the recent " Severe Accident Risks: An Assessment for i 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 improved 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 assuring that radiological doses from normal operation and postulated accidents will be l 5
._.- n acceptably low, that natural phenomena and potential man-made hazards will be appropriately accounted for in the design of the plant, and that site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant. The Commission has also had a long standing policy of siting reactors away from .
densely populated centers, and is continuing this policy in this rule. The Commission is incorporating basic reactor site criteria in this rule to accomplish the above purposes. The Commission is retaining source term 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 performance rather than siting. No specific source term is referenced in Part 50. Rather, the source term is required to be one that is "... assumed 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 terms. '
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 3 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 compensate for otherwise poor site >
conditions. i 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 immediate vicinity of the plant has- been a requirement for siting power reactors from the very 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 from potential man-related hazards. The Commission considers an exclusion area to be an essential feature of a reactor site and is retaining this requirement, in Part 50, 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.
6
The present regulation requires that the exclusion area be of such size that an individual located at any point on its boundary for two hours immediately 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 rem to the thyroid gland. A footnote in the present regulation 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 Commission'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 in order to provide assurance of low risk to the public under postulated accidents. The Commission, based upon extensive experience in applying this criterion, and in recognition of the conservatism of the assumptions in its application (a large fission pr'oduct 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 meteorological 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 estimated to be about 0.1 rem. 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 j 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 Commission's regulations and i
was also based upon two considerations. First, since it utilizes a risk I ' consistent methodology to assess the radiological impact of all relevant nuclides upon all body organs, use of TEDE promotes a uniformity and 3 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 4 nuclides, besides the noble gases and iodine are predicted to be released into containment. , The Commission 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 (QH0s) for it have been established in the Commission's Safety Goal policy. Although the supplementary information 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. 7
4 4 However, the Commission specifically requested comments on whether the current dose crf teria should be modified to utilize the total effective dose equivalent or TIDE 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 dne criterion should also include a " capping" limitation, that is, an additional requirement that the dose to any inoividual organ not be in excess of some fraction of the total. Based on the comments 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 300 rem to the thyroid gland. These commenters noted that the use of organ weighting factors of I 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 res, 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 Commission has decided to adopt a value of 25 rem TEDE as the dose acceptance criterion for the final rule. The bases for this decision follows. First, the Commission 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 permits 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 Commission 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 emergency 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 licensing reviews, and that all operating plants would be able to meet a dose criterion of 25 rem TEDE. Hence, the Commission concludes that, in practice, use of the organ weighting factors of Part 20 together with a dose criterion of 25 rem TEDE, 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" value that bears no relation to what might be experienced by an actual person in an accident. Although the Commission 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 8
boundary. While the duration of the time period remains a't a value of two , hours, the proposed rule stated that this time period not be fixed in regard l to the appearance of fission products within containment, but that various j two-hour periods be examined with the objective that the dose to an individual e not be in excess of 25 rem TEDE for any two-hour period after the appearance i of fission products within containment. The Commission proposed this change to ! reflect-improved understanding of fission product release into the containment under severe accident conditions. For an assumed instantaneous release of l fission products, as contemplated by the present rule, the two hour period l- that commences 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 commencing with the onset j 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
- Commission proposed that various two-hour periods be examined to assure that the dose t'o 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. 1 A number of comments teceived in regard to this proposed criterion i stated that so-called " sliding" two-hour window for dose evaluation at the l exclusion area boundary was confusing, illogical, and inappropriate. Several ,
commenters 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 comment expressed the view that a dose evaluated for a " sliding" 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
" maximum" two hour value was being ignored.
Although the Commission 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 s'nould not be taken too literally with regard to the I 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 mitigating postulated accidents. For these reasons, the Commission 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). 1 B. ELte Discersion Factors Site dispersion factors have been utilized to provido an acassment of dose to an individual as a result of a postulated I accident. Since the Commission is requiring that a verification be made that 9 i i a ne
._ m _ . , _ .
. o ,
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 j 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. l C. Low Population Zone. The pmsent regulation requires that a low !
population zone (LPZ) be defined i!:weliately beyond the exc?usion area. Residents are permittad in this area, but the number and density must be such that there is a reasonable probability that appropriate protective measures I could be taken in their behalf in the event of a serious accident. In i addition, the nearest densely populated center contaiMn; i., ore than about 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 must not be in excess of the dose. values 'given in the regulation. While the Commission 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 Commission 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 Commission 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 l excess of 25 rem TEDE. l
\
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 Milita.rv Fap_ililin As for natural phenomena, it has been a long-standing NRC staff practice to review man-related activities in the site vicinity to provide assuraace that potential hazards associated with such facilities or transpor'stico routes will pose no undue risk to any plant proposed to be located at the site. The l final rule codifies this practice. F. Adeauacy of Security Plans The rule requires that the . characteristics of the site be such that adequate security plans and measures I for the plant could be developed. The Commission envisions that this will entail a small secure area considerably smaller than that envisioned for the exclusion area. 10
G. Adeauacy of Emeraency Plans The rule also. requires that the site characteristics be such that adequate plans to carry out protective measures for members of the public in the event of emergency could be developed. H. Sitina Away From Densely Poculated Centers Population density considerations beyond the exclusion area have been required since issuance of Part 100 in 1962. The current rule requires a " low population zone" (LPZ) beyond the immediate exclusion area. The LPZ boundary must be of such a size that an individual located at its outer boundary must not receive a dose in excess of the values given 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 representing an upper. limit are conceivable, although idghly improbable, it was considered desirable to provide for protection against excessive exposure doses to people in large centers, where effectin 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 Commission'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 usaful insights for considering 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 release" (without defining this precisely). Based upon NUREG-1150, the 11 c -I
i probability of a core-melt accident together with early containment failure or byp' ass for some current generation LWRs is estimated to be between 10~5 and 10' per reactor year. For future plants, this value is expected to be less L than 10
- per reactor year.
s If a reactor was located nearer to a large city than current NRC 1 practice permitted, 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 10-' per reactor year or lower e have been regarded in past licensing actions to be " incredible", and as such, i have not been required to be incorporated into the design basis of the plant. i 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 i
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, i 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 I
- 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 j natural processes of fallout and rainout before reaching the city. Analyses !
2 indicate that if a reactor were located at distances ranging from 10 to about ! i 20 miles away from a city, depending upon its size, the likelihood of exposure I j of large numbers of people within the city would be reduced by factors of ten ! 3 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 j characteristics sufficiently low that the safety of the public is reasonably ;
assured by the reactor and plant design and operation itself, resulting in a very low likelihood of occurrence of a severe accident. Such a plant can 1 i satisfy the QH0s of the Safety Goal with a very small exclusion area distance 4 (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 l l or less. With regard to population density beyond the exclusion area, siting j ! a reactor closer to a densely populated city than is current NRC practice j i would pose a very low risk to the populace. i Nevertheless, the Commission concludes that defense-in-depth ! l considerations and the edditional enhancement in safety to be gained by siting i i reactors away from sensely populated centers should be maintained. 1 j The Commisslon is incorporating a two-tier approach with regard to
, population density and reactor sites. The rule requires that reactor sites be j located away from very densely populated centers, and that areas of low
' population density are, generally, preferred. The Commission believes that a j site not falling within these two categories, although not preferred, can be found acceptable under certain conditions. i The Commission is not establishing specific numerical criteria for ;
! evaluation of population density in siting future reactor facilities because e the acceptability of a specific site from the standpoint of population density
! must be considered in the overall context of safety and environmental considerations. The Commission's intent is to assure that a site that has l 12 1
E a
n . i i significant safety, environmental or economic advantages is not rejected l 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 acceptability. Thus, it must 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 Commission 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 i nuclear project and other reactors be sited at least 40 miles from major population centers." The petitioner also stated that " locating reactors in i sparsely-populated areas ...has been endorsed in non-binding NRC guidelines . for reactor siting." The petitioner did not specify what constituted a major i population center. The only NRC guidelines concerning population density in i 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 from the reactor, not 40 miles. Regulatory Guide 4.7 does provide effective separation from population i 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. i The Commission has examined these guidelines with regard to the Safety . Goal. The Safety Goal quantitative health objective in regard to latent l cancer fatality states that, within a distance of ten miles (16 km) from the i reactor, the risk to the population of latent cancer fatality from nuclear i 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 Commission has also t 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 i analysis done in support of this proposed change in regulation, the likelihood of permanent relocation of people located more than about 20 miles (32 km) ; from the reactor as a result of land contamination from a severe accident is j very low. A revisior, of Regulatory Guide 4.7 which incorporated this finding that population density guidance beyond 20 miles was not needed in the evaluation of poterMal reactor sites was issued for comment at the t6e of i the proposed rule. N comments were' received on this aspect of the guide. ! Therefore, the Commission concludes that the NRC staff guidance in i Regulatory Guide 4.7 provide a means of locating reactors away from population l' centers, including " major" population centers, depending upon their size, that t 13 l l l
< . 9,
. l would limit societal consequences significantly, in the ev'ent of a severe accident. The Commission 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 Commission also believes that a higher population density site could be found to be acceptable, compared to a lower l 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 l Commission has decided not to adopt the proposal by Free Environment, l Incorporated.
l The Commission also notes that future population growth around a nuclear l power plant site, as in other areas of the region, is expected but cannot be l predicted with great accuracy, particularly in the long-term. Population
- l. growth in the site vicinity will be periodically factored into the emergency l plan for the site, but since higher population density sites are not unacceptable, per se, the Commission does not intend to consider license conditions or 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 Commission wishes to emphasize that population considerations as well as other siting 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 l 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. These requirements apply to future applicants for a construction permit, design certification, or an operating license. The Commission will consider ! l 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. ; B. Seismic 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 l reflect new information and research results, and incorporate the intentions i of this regulatory action as defined in Section III of this rule. Much of the following discussion remains unchanged from that issued for public comment (59 FR 52255) because there were no comments which necessitated a major change to the regulations and supporting documentation.
- 1. Separate Sitina from Desion.
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 i 14 1
g . _ . _ _ _ . _ . _ _ _ _ _ . _ . _ l
~
l f in 10 CFR Part 50. Because the revised criteria presented in the regulation , j will not be applied to existing plants, the licensing basis for existing i nuclear power plants must remain part of the regulations. The criteria on i seismic and geologic siting would be designated as a new s 100.23 to Subpart B 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 Reaulation.
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 l 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 terms 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 i based on limited geophysical and geological information and will be revised as necessary when more 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 i 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 l analyses (for instance, probabilistic) in the licensing process. l The final regulation is streamlined, becoming a new section in Subpart B l 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 determine the geological, seismological, and engineering characteristics of the proposed site, and (c) requirements to determine the Safe Shutdown Earthquake Ground Motion (SSE), to determine the potential for surface deformation, 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 determine 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 Determination of Safe Shutdown Earthquake Ground Motions." l
- 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 i 15
y 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 to be suitably conservative, the approach has not explicitly recognized uncertainties in geosciences 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 "probabilistic" re.Lhods have been designed to allow explicit ine 'rporation 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 models 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, :md the understanding of tectonics. Therefore, methods other than the probabilistic methods, such as sensitivity analyses, may be adequate for some sites to account for uncertainties. Methods accepte19 to the NRC staff for implementing the regulation are described in Regulat;.y 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, Determine 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. 16
- ""% v
. . 1 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) 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 i as structural geology, stratigraphy, and topography and to account for site-specific geotechnical properties in establishing the design basis ground motion. 'In order to incorporate local site factors and advances in ground l l motion attenuation models, ground motion characteristics are determined using l l the procedures outlined in Standard Review Plan Section 2.5.2, " Vibratory l Ground Motion," Revision 3. l The NRC staff's review approach to evaluate ground motion estimates is l described"in SRP Section 2.5.2, Revision 3. This review takes into account the information 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 l 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 Earthauake.
The existing regulation (10 CFR Part 100, Appendix A, Section V(a)(1)(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 determined ..." The location of the seismic input motion 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 moves 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 motion in the free-field. The final regulation is also consistent with the resoiution of Unresolved Safety Issue (USI) A-40, " Seismic Design Criteria" (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 , l ground acceleration of at least 0.19 l
- 5. Value of the Ooeratina Basis Earthauake Ground Motion (OBE) and ,
Reauired OBE Analyses. 17 n , - . . . . - ,. , , , . . , , - , - , . . ,
, _7 The existing regulation (10 CFR Part 100, Appendix A, Section V(a)(2)) '
states that the maximum vibratory ground motion 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, Section VI(a)(2)) states that the engineering method used to insure that' ; structures, systems, and components 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 the 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 Commission Piping Review Committee," Vol.5, April 1985, (Table 10.1) 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 l Requirements," the NRC staff states that it agrees that the OBE should not l control the design of safety systems. Furthermore, the final safety i 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 was set so that the vibratory ground motion should not induce stresses exceeding the allowable stress limits originally required for the OBE design. The final regulation allows the value of the OBE to be set at (1) one-third or less of the SSE, where OBE requirements are satisfied without an explicit response or design analyses being performed, 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: l first, plant shutdown is required if vibratory ground motion exceeding that of ) the OBE occurs (discussed below in Item 6, Required Plant Shutdown), and 1 second, the amount of analyses associated with the OBE. An applicant may determine that at one-third of the SSE level, the probability of exceeding the OBE vibratory ground motion 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 performed 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 structurcs, 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, 18
. . i l
strain and deformation limits when subjected to the effects of the OBE in ; combination with normal operating loads. As stated, it is determined that if an OBE of one-third or less of the l SSE is used, the requirements of the OBE can be satisfied without the ; applicant performing any explicit response analyses. 'In this case, the OBE l serves the function of an inspection and sh~utdown earthquake. Some minimal design checks and the applicability of this position to seismic base isolation of buildings are discus:ed below. There is high confidence that, at this ground-motion 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 margins-type assessment of seismic events. A PRA-based seismic , margins analysis will consider sequence-level High Confidence, Low Probability j of Failures (HCLPFs) and fragilities for all sequences leading to core damage or containment 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 Analysis of External Events, memoranduni from 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 Pl ants") . 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. _ In SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining to Evolutionsry and Advance Light-Water Reactor (ALWR) Designs," the NRC staff requested Commission 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 components 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 OBE, but not the SSE. Also, the NRC staff has evaluated the effect on safety of eliminating the OBE from the design load combinations for selected structures, systems, and components and 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 3 all earthquake levels. Decisions pertaining to the response analyses ! associated with base isoiated facilities will be handled on a casa by case ! basis. i
- 6. Reauired Plant Shutdown.
i l 19
i 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 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 $50.36(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 Commission 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 Commission 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 l 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 l specifications. 1 The final regulation treats plant shutdown associated with vibratory l 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 l require a process leading to plant shutdown for licensees of nuclear power l plants that comply 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. Immediate shutdown could be required until.it is determined that structures, systems, and components needed for safe shutdown are still functional. Regulatory Guide 1.166, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions," provides guidance acceptable to the NRC staff for determinir.g 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 'I 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 Commission 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. (This unavailability may be due to earthquake ! related damage.) 20 i I 1
., + w .w , w--- -
, ~ Regulatory Guide 1.167, " Restart of a Nuclear Power P'lant 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 structures prior to plant restart. This guidance is also based on EPRI reports. Prior to resuming operations, the licensee must demonstrate to the Commission 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 documented in written reports and submitted to the Director, Office of Nuclear Reactor Regulation. The licensee shall not resume operation until authorized to do so by the Director, Office of Nuclear Reactor Regulation.
- 7. Clarify interoretations.
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 VI(a). With the relocation of the earthquake engineering criteria to Appendix S to Part 50 and the relocation and modification to dose guidelines in 550.34(a)(1), 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. VI. Related Regulatory Guides and Standard Review Plan Sections l The NRC is developing the following regulatory guides and standard I review plan sections to provide prospective licensees with the necessary guidance for implementing 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, " Identification and Characterization of Seismic Sources and Determination of Shutdown Earthquake Ground Motions." The g;;ide provides general guidance and recommendations, describes acceptable procedures and provides a list of references that present acceptable rathodologies to identify and characterize capable tectonic sources and seismogenic sources. Section V.B.3 of this rule describes the key elements.
- 2. Regulatory Guide 1.12, Revision 2, " Nuclear Power Plant Instrumentation for Earthquakes." The guide describes seismic instrumentation type end location, operability, characteristics, installation, actuation, and maintemnce that J acceptable to the NRC staff. ,
- 3. Regulator 2 Guide 1.166, " Pre-Earthquake Planning and Immediate '
Nuclear Power P17.at Operato" Post-Earthquake Actions." The guide provides guidelines that are eniitable to the NRC staff for a timely evaluation of the -1 recorded seismic instrumentation data and to determine whether or not plant I shutdown is required. l 21 j
. v--
1
- 4. Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Dcwn by a Seismic Event." The guide provides guidelines that are acceptable to the NRC staff for performing inspections and tests of nuclear power plant equipment and structures prior to restart of a plant that has been shut dcwn because of a seismic event.
- 5. Standard Review Plan Section 2.5.1, Revision 3, " Basic Geologic and 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 will be issued as final guides without public comment subsequent to the publication of the final regulations. The following regulatory guides will be revised to incorporate editorial I 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. l
- 1. 1.57, " Design Limit: and Loading Combinations for Metal Primary Reactor Containment S,vstem Components."
- 2. 1.59, " Design 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, " Bases for Plugging Degraded PWR Steam Generator Tubes."
- 8. 1.122, " Development of Floor Design Response Spectra for Seismic '
Design of Floon-Supported Equipment or Components." The following regulatory guides will be revised to update the design or analysis philosophy, for example, to change OBE to a fraction of the SSE:
)
22 i _ _ _ . _ ., - __ _ _ - _ _ _ _ -m - - .--
- o. .
- 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."
- 4. 1.100, " Seismic Qualification of Electric and Mechanical Equipment l for Nuclear Power Plants."
i
- 5. 1.124, " Service Limits and Loading Combinations for Class 1 i
Linea stype Component Supports." l 6. 1.130, " Service Limits and Loading Combinations for Class 1 Plate-ancbShell-Type Component Supports." [
- 7. 1.132, " Site Investigations for Foundations of Nuclear Power Plants."
- 8. 1.138, " Laboratory Investigations of Soils for Engineering i
Analysis and Design of Nuclear Power Plants." 4- .9. l.142, " Safety-Related Concrete Structures for Nuclear Power
' Plants (Other than Reactor Vessels and Containments)."
- 10. 1.143, " Design Guidance for Radioactive Waste Management Systems, Structures, and Components Installed in Light-Water-Cooled Nuclear Power Plants."
Minor and conforming 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 comment 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. Copies of NUREG-0625, NUREG-1061, NUREG-1150, NUREG-1451, NUREG-1462, NUREG-1503, and NUREG/CR-2239 may be purchased from the Superintendent of Documents, U.S. Government Printing Office, Mail Stop SSOP, Washington, DC 20402-9328. ' Copies also are available from the National Technical Information 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 from the Government Printing Office (GP0) at the current GP0 price. Information on current GP0 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 Information 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. 23
4 W
~
IX. Summary of Comments on the Proposed Regu1ations. l A. Reactor Siting Criteria (Honseismic). J Eight organizations or individuals commented on the nonseismic aspects 1 of the second proposed revision. The fi si, proposed revision issued for comment in October 20, 1992, (57 FR 47802) elicited strong comments in regard to proposed numerical values of population density and a minimum distance to l l the exclusion area boundary (EAB) in the rule. The second proposed revision l (October 17,1994; 59 FR 52255) would delete these from the rule by providing 1 guidance on population density in a Regulatory Guide and determining the I
; distance to the EAB and LPZ by use of source term and dose calculations. The l j rule would contain basic site criteria, without any numerical values. l
- Several commentors representing the nuclear industry and international
5 nuclear organizations stated that the second proposed revision was a ! significarit improvement over the first. proposed revision, while the only public interest group commented that the NRC had retreated from decoupling siting and design in response to the comments of foreign entities. _< Most comments 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 period, and the question of whether an organ capping dose should be
- adopted.
Virtually all commenters 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 commenters 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 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 commenters opposed the use of an organ capping dose. The only public interest group that commented did not object to the use of TEDE, favored the proposed dose value of 25 rem, and supported an organ capping dose.
B. Seismic and Earthquake Engineering Criteria. Seven letters were received addressing either the regulations or both the regulations and the draft guidance documents identified in Section VI (except DG-4003). An additional five letters were received addressing only the guidance documents, for a total of twelve comment letters. A document,
" Resolution of Public Comments on the Proposed Seismic and Earthquake Engineering Criteria for Nuclear Power Plants," is available explaining the NRC's disposition of the comments received on tha regulations. A copy of this document has been placed in the NRC Public Docunent Room, 2120 L Street NW.
(Lower Level), Washington, DC. Single copies i.re available from Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory 24
_ _ ._ _ _ . . _ __ _ _ . _ ~.-._ _ _ _ _ _ _ . _ _ ii . i-1 Commission, Washington, DC 20555-0001, telephone (301) 415-6010. A second
- document, " Resolution of Public Comments on Draft Regulatory Guides and i Standard Review Plan Sections Pertaining to the Proposed Seismic and i Earthquake Engineering Criteria for Nuclear Power Plants," will explain the L NRC's disposition of the comments received on the guidance documents. The l Federal Register notice announcing the avaliability of the guidance documents will also discuss how to obtain copies of the comment resolution document.
- A summary of the major comments on the proposed regulations follows.
l Supplementary Information , i 2 Section III, Genesis (Application) Comment: The Department of Energy (Office of Civilian Radioactive Waste
- Management), requests an explicit statement on whether or not s 100.23 applies l
. to the Mined Geologic Disposal System (MGDS) and a Monitored Retrievable
- Storage (MRS) facility. The NRC has noted in NUREG-1451, " Staff Technical Position o'n 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 l contemplated revisions to Part 100 would 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. j 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 I two other parts of the regulation. They are (1) Part 40, " Domestic Licensing j of Source Material," Appendix A, " Criteria Relating to the Operation of j 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 l 1 72, " Licensing Requirements for the Independent Storage of Spent Nuclear Fuel j and High-Level Radioactive Waste," Paragraphs (a)(2), (b) and (f)(1) of 572.102.
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 i 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 l through a rulemaking, revision of NUREG-1451, or other appropriate mechanism. ~ Section V(B)(5), "Value of the Operating Basis Earthquake Ground Motion (0BE) and Required OBE Analysis." i j Comment: One commenter, ABB Combustion Engineering Nuclear Systems, i 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 l
comments, from G.C. Slagis Associates, stated that the proposed rule in the area of required OBE analysis is not sound, not technically justified, and not j appropriate for the design of pressure-retaining components. The following )' 25 4 J. J
. o are specific comments (limited to the design of pressure-retaining components to the ASME Boiler and Pressure Vessel Section III rules) that pertain to the supplemental information to the proposed regulations, item V(B)(5), "Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analysis."
(1) Comment: Disagrees with the statement in SECY-79-300 that design for a single limiting event and inspection and evaluation for earthquakes in excess of some specified limit may be the most sound regulatory approach. It is not feasible to inspect for cyclic damage to all the pressure-retaining components. Visually inspecting for permanent deformation, or leakage, or failed component supports is certainly not adequate to determine cyclic damage. Respanse: The NRC agrees. Postearthquake inspection and evaluation guidance is described in . Regulatory Guide 1.167 (Draft was DG-1035), " Restart of a Nuclear Power Plant Shut Down by an Seismic Event." The guidance is not limited to visual inspections; it includes inspections, tests, and analyses including fatigue analysis. . (2) Comment: Disagrees with the NRC statement in SECY-090-016 that the OBE should not control design. There is a problem with the present requirements. Requiring design for five OBE events at one-half SSE is unrealistic for most (all?) sites and requires an excessive and unnecessary number of seismic supports. The solution is to properly define 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 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) Comment: It is not technically justified to assume that Section III components will remain 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 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 component 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 dominates the load combination. 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. 26
i Eliminating the OBE will, however, directly affect the current methods used to evaluate the adequacy. of cyclic and secondary stress effects in the l 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 i to equipment and structures at various elevations because both of these ' effects are currently evaluated only for OBE loadings. Accordingly, to l account for earthquake cycles in the fatigue analysis of piping systems, the staff proposes to develop guidelines for selecting a number of SSE cycles at a fraction of the peak amplitude 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 OBE were proposed in i SECY-93-087. Commission approval is documented 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) Comment: There is one major flaw in the "SSE only" design approach. The equipment designed for SSE is limited to the equipment necessary to assure j 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 "necessary for continued operation without undue risk to the health and safety of the public." Hence, f by this rule, it is possible that some equipment necessary for continued ; operation will not be designed for SSE or OBE effects. Response: The NRC does not agree that the design approach is flawed. It is not possible that some equipment necessary for continued aft operation will not be designed for SSE or OBE effects. General Design Criterion 2, ,
" Design Bases for Protection Against Natural Phenomena," of Appendix A,
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50 requires l 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 l structures, systems,.and components important to safety to withstand the l effects of earthquakes. Regulatory Guide 1.29, " Seismic Design i Classification," describes a method acceptable to the NRC for identifying and J classifying those features of light-water-cooled nuclear power plants that l 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 l 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 comments that were provided on the earlier proposed rulemaking package (October 20, 1992;.57 FR 47802) and considered that the seismic portion of the proposed rulemaking package is nearing maturity and with the inclusion of industry's 27
. . - - - - -- . ._. . .- - - . . ~ - .
i j comments (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 i 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 2 deterministic and probabilistic evaluations must be conducted to determine 1 site suitability and seismic design requirements for the site. [ Note: the j commenters do not agree with the NRC staff's deterministic 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 i not support the NRC staff's deterministic check of the applicants submittal j (SRP Section 2.5.2). These items are addressed in the document pertaining to j
- comment resolution of the draft regulatory guides and standard review plan I
- sections.]
t Comment: NEI, Westinghouse Electric Corporation, and Yankee Atomic Electric Corporation recommend 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 4
given confirmatory foundations evaluations [ Regulatory Guide 1.132, but not the geologic, geophysical, seismological investigations in Regulatory Guide }, 1.165]. Response: The NRC has determined that the use of_ a spectral shape l anchored to 0.3g peak ground acceleration as a standardized design level would i be appropriate for existing central and eastern U.S. sites based on the
- current state of knowledge. However, as new information 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 i to make that assessment. Therefore, it is not appropriate to codify the j request.
! Comment: NEI racommended a rewording of Paragraph (a), Applicability. Although unlikely, an applicant for an operating license already holding a
- construction permit may elect to apply the amended methodology and criteria in
- Subpart 8 to Part 100.
- Response: The NRC will address this. request on a case-by-case basis j rather than through a generic change to the regulations. This situation l
- pertains to a limited number of facilities in various stages of construction. 1
, Some of the issues that must be addressed by the applicant and NRC during the ; i operating license review include differences between the design bases derived '
- from the current and amended regulations (Appendix A to Part 100 and s 100.23, l respectively), and earthquake engineering criteria such as, OBE design ,
requirements and OBE shutdown requirements. ~ j Appendix S to 10 CFR Part 50
- Support for the NRC position perte.ining to the elimination of the
- Operating Basis Earthquake Ground Motion (OBE) response analyses has been 4
? 28
documented in various NRC publications such as SECY-79-300, SECY-90-016, SECY- ' 93-087, and NUREG-1061. 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. In addition, similar activities are being done in foreign countries, for instance, Germany. (Additional discussion is provided in Section V(B)(5) of this rule). 4 Comment: The American Society of Civil Engineers (ASCE) recommended that the seismic design and engineering criteria of ASCE Standard 4, " Seismic Analysis of Safety-Related Nuclear Structures and Commentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures," be incorporated by reference into Appendix S to 10 CFR Part 50. Response: The Commission has determined that new regulations will be , more streamlined 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 4 inhibit the use of needed latitude in judgement. Therefore, it is common NRC 2 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 in regulatory guides and standard review plan sections. ASCE Standard 4 is cited in the 1989 revision of Standard Review Plan Sections
- l. 3.7.1, 3.7.2, and 3.7.3.
j Comment: The Department of Energy stated that the required consideration i of aftershocks in Paragraph IV(B), Surface Deformation, is confusing and recommended 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 VI(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 !
recommended change will make Paragraphs IV(a)(1), " Safe Shutdown Earthquake i Ground Motion," and IV(b), " Surface Deformation, of Appendix S to 10 CFR Part 50 consistent. , a i X. Small Business Regulatory Enforcement Fairness Act In accordance with the Small Business Regulatory Enforcement Fairness l Act of 1996 the NRC has determined that this action is not a major rule and ; has verified this determination with the Office of Information and Regulatory . l Affairs of OMB. XI. Finding of No Significant Environmental Impact: Availability T 4 The Commission has determined under the National Environmental Policy Act of 1969, as amended, and the Commission'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. 29
-, , , - - . ,- - -+
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 Commission 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 the public health and safety over the current regulations. The amendments do not affect nonradiological plant effluents and have no other environmental impact. The addition of s100.23 to 10 CFR Part 100, and the addition of Appendix 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 maintenance of seismic instrumentation. 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 instrumentation ensures its operability during earthquakes. The location of the seismic instrumentation is similar to that in the existing nuclear power plants. The amendments 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 Document Room, 2120 L Street NW. (Lower Level), Washington, DC. Single copies of the environmental assessment and finding of no significant impact are available from Mr. Leonard Soffer, Office of the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1722, or Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, 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.). These requirements were approved 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 hours per response, including the time for reviewing instructions, searching existing data sources, gathering and maintaining the data needed, and completing and reviewing the collection of information. Send comments 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 BJS19NRC. GOV; and to the Desk Officer, Office of Information and Regulatory Affairs, NE08-10202, (3150-0011 and 3150-0093), Office of Management and Budget, Washington, DC 20503. Public Protection Notification 30
O
- The NRC may not conduct or sponsor, and a person is not required to respond to, a collection of information unless it displays a currently valid OMB control number.
XIII. Regulatory Analysis The Commission has. prepared a regulatory analysis on this regulation.
- The analysis examines the costs and benefits of the alternatives considered by the Commission. Interested persons may 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 Mr. Leonard Soffer, Office of the Executive Director for Operations, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-1722, or i Dr. Andrew J. Murphy, Office of Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, telephone (301) 415-6010. l XIV. Regulatory Flexibility Certification As required by the Regulatory Flexibility Act of 1980, 5 U.S.C. 605(b), the Commission certifies that this regulation does not have a significant economic impact on a substantial number of small entities. This regulation t affects only the licensing and operation of nuclear power plants. The l companies that own these plants do not fall within the definition of "small entities" set forth in the Regulatory Flexibility Act or the size standards established by the NRC (April 11, 1995; 60 FR 18344). XV. Backfit 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)(1). 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 10 CFR Part 50 - Antitrust, Classified information, Criminal penalties, Fire protection, Intergovernmental 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, Limited 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. 31
n 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, as amended, and 5 U.S.C. 552 and 553, the NRC is adopting the following amendments to 10 CFR Parts 50, 52, and 100.
PART 50 - DOMESTIC LICENSING OF PRODUCTION AND UTILIZATION FACILITIES
- 1. The authority citation for Part 50 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 amended (42 U.S.C. 2132, 2133, 2134, 2135, 2201, 2232, 2233, 2236, 2239, 2282); sec's. 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 50.10 also issued under secs. 101, 185, 68 Stat. 955 as amended (42 U.S.C. 2131, 2235), sec. 102, Pub. L. 91-190, 83 Stat. 853 (42 U.S.C. 4332). Sections 50.13, 50.54(dd) and 50.103 also issued under sec. 108, 68 Stat. 939, as amended (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. 97-415, 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 -- 50.81 also issued under sec.184, 68 Stat. 954, as amended (42 U.S.C. 2234). Appendix F also issued under sec. 187, 68 Stat. 955 (42 U.S.C. 2237).
- 2. Section 50.2 is revised by adding in alphabetical order the definitions for Committed dose eauivalent, Committed effective dose eauivalent, Deep-dose eauivalent, Exclusion area, Low coDulation Zone, SafetV-related structures. systems. and components and Total effective dose eauivalent to read as follows s 50.2 Definitions. * * * *
- j Committed dose eauivalent means the dose equivalent to organs or ;
tissues of reference that will be received from as intake of radioactive ! material by an individual during the 50-year perica following the intake. 1 Committed effective dose eauivalent is the sum of the products of I the weighting factors applicable to each of the body organs or tissues that l are irradiated and the committed dose equivalent to these crgans or tissues. l Deeo-dose eauivalent, which applies to external whole-body exposure, is the dose equivalent at a tissue depth of I cm (1000mg/cm ), i
* * * *
- 1 32
Exclusion 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 normally 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 copulation zone means the area immediately surrounding the exclusion area which contains residents, the total number and density of which are such t, hat 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. 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 shutdown 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 in s 50.34(a)(1) or,s 100.11 of this chapter.
Total effective dose eauivalent (TEDE) means the sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).
- 3. 1.n s50.8, paragraph (b) is revised to read as follows:
5 50.8 Information collection requirements: DM8 approval. (b) The approved information collection requirements contained in this part appear in ss50.30, 50.33, 50.33a, 50.34, 50.34a, 50.35, 50.36, 50.36a, 33
_ ~ ._- . _ _ _ _ 1 4 50.48, 50.49, 50.54, 50.55, 50.55a, 50.59, 50.60, 50.61, 5'0.63, 50.64, 50.65, 50.71, 50.72, 50.80, 50.82, 50.90, 50.91, and Appendices A, B, E, G, H, I, J, K, M, N, 0, Q, R, and S. - 1
- 4. In s50.34, footnotes 6, 7, and 8 are redesignated as footnotes 8, I 9 and 10 and paragraph (a)(1) is revised and paragraphs (a)(12), ;
(b)(10), and (b)(11) are added to read as follows: ' l s 50.34 Contents of applications; technical information. ; l (a) * * * (1) Stationary power reactor applicants for a construction permit l 7 pursuar,t to this part, or a design certification or combined license pursuant ' a to Part 52 of this chapter who apply on or after [ INSERT EFFECTIVE DATE OF THE
- FINAL RULE], shall comply with paragraph (a)(1)(ii) of this section. All other :
applicants for a construction permit pursuant to this part or a design I certification or combined license pursuant to Part 52 of this chapter, shall comply with paragraph (a)(1)(1) of this section. l (i) A description and ~ safety assessment of the site on which the-- ) facility is to be located, with appropriate attention to features affecting 1 facility design. Special attention should be directed to the site evaluation ! factors identified in Part 100 of this chapter. The 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 4
site evaluation factors identified in Part 100 of this chapter, assuming that the facility will be operated at the ultimate power level which is contemplated 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 l
radioactive fission products. The following power reactor design. characteristics and proposed operation will be taken into consideration by the l Commission: i (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; i
34 I
m - (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 release of radioactive material to the environment can occur. Special attention must be directed to plant design features intended to mitigate the l radiological consequences of accidents. In performing this assessment, an j applicant shall assume a fission product release
- from the core into the containment assuming that the facility is operated at the ultimate power level contemplated. The applicant shall perform an evaluation and analysis of the postulated fission product release, using the expected demonstrable containment leak rate and any fission product cleanup systems intended to ;
mitigate the consequences of the accidents, together with applicable site j characteristics, including site meteorology, to evaluate the offsite 1 radiological consequences. Site characteristics must comply with Part 100 of l this chapter. The evaluation must determine that: j (1) An individual located at any point on the boundary of the exclusion area for any 2 hour period following the onset of the postulated ; fission product release, would not receive a radiation dose in excess of 25 : ! rem' tota 1' effective dose equivalent (TEDE). (2) An individual located at any point on the outer boundary of i the low population zone, who is exposed to the radioactive cloud resulting ! l from the postulated fission product release (during the entire period of its l passage) would not receive a radiation dose in excess of 25 rem total l effective dose equivalent (TEDE). (E) 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. (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 comply with the earthquake engineering criteria in Appendix S to this part.
- The fission product release assumed for this evaluation should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events.
Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release into the containment of appreciable quantities of fission products.
' A whole body dose of 25 rem has been stated to correspond nianerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP reconnendations at the time could be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5,1959). However, its use is not intended to imply that this nisaber constitutes an acceptable limit for an smargency 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 features with respect to postulated reactor accidents, in order to assure that such designs provide assurance of low risk of public exposure to radiation, in the event of such accidents.
35
. s--
(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 comply 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 OF THE FINAL RULE), the earthquake 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 combined license pursuant to Part 52 of this chapter, shall provide a description and safety assessment of the site and of the facility as in s50.34(a)(1)(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. l
- 5. In s50.54, paragraph (ff) is added to read as follows:
l l s50.54 Conditions of licenses. (ff) For licensees of nuclear power plants that have implemented the 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 i operations, the licensee shall demonstrate to the Commission that no j 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 maintained.
- 6. 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 36
O e 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 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 combined license is required by 550.34(a)(12), (b)(10), 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 in s 50.54(ff), nuclear power plants that have implemented the earthquake engineering criteria described herein must shut down if the criteria in Paragraph IV(a)(3) of this appendix are exceeded. Thes'e criteria implement 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 appendix are within the scope of investigations permitted by s50.10(c)(1). III. Definitions As used in these criteria: Combined 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. Desian Certification means a Commission 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." The Operatina Basis Earthauake Ground Motion (OBE) 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 j is only associated with plant shutdown and inspection unless specifically selected by the applicant as a design input. . A resoonse 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 j 37
_ _ .~. _-_._ _ _ _ _ _ _ _
. m W
, damping value. The response spectrum is calculated for a specified vibratory
- motion input at the oscillators' supports.
The Safe Shutdown Earthauake Ground Motion (SSE) is the vibratory ground motion for which certain structures, systems, and components ;ust be designed i to remain functional. The structures. systems. and components reauired to withstand the !
- effects of the Safe Shutdown Earthauake Ground Motion er surface deformatun '
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 l (3) The capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures comparable to the guideline exposures of s50.34(a)(1)(ii). 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. l IV. Application To Engineering Design l The following are pur:uant to the seismic and geologic design basis j requirements of s100.23 of this chapter: 1 (a) 'Jibratory Ground Motion. i (1) Safe Shvi.6:r. Drthauake Ground Motion. The Safe Shutdown Earthquake Ground A ' otion must b3 characterized by free-field ground motion response spectra at the free ground surface. In view of the limited data available on vibratory ground motions 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 q free-field at the foundation level of the structures must be.an appropriate response spectrum with a peak ground acceleration of at least 0.1g. The nuclear power plant must be designed so that, if the Safe Shutdown Earthquake Ground Motion occurs, certain structures, systems, and components will remain functional and within applicable stress, strain, and deformation 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
)f the nuclear power plant must also take into account the possible effects of the Safe Shutdown Earthquake Ground Motion on the facility foundations by ground disruption, such as' fissuring, lateral spreads, differential settlement, liquefaction, and landsliding, as required in s100.23 of this crepter. ,
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. 38
U 8 The evaluation must 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, and 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 Ground Motion. (1) 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: i (A) One-third or less of the Safe Shutdown Earthquake Ground Motion design response spectra. The requirements associated with this Operating Basis Earthquake Ground Motion in Paragraph (a)(2)(1)(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 Mot, ion design response spectra. Analysis and design must be performed , to demonstrate that the requirements associated with this Operating Basis
' Earthquake Ground Motion in Paragraph (a)(2)(1)(B)(I) are satisfied. The design must 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 components of the nuclear power plant necessary for continued operation without undue risk to the health and safety of the public must 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 tround Motion or if significant plant damage ; occurs, the licensee must shut dow's 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 i Ground Motion, the licensee iiust consult with the Commission and must propose a plan-for the timely, safe rhutdown of the nuclear power plant. Prior to resuming operations, the licensee must demonstrate to the Commission that no functional damage has occurred to those features necessary for continued '
operation without undue risk to the health and safety of the public. (4) Required Seismic Instrumentation. Suitable instrumentation must be provided so that the seismic response of nuclear power plant features , l important to safety can be evaluated promptly after an earthquake. - (b) Surface Deformation. 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, systems, and components will remain functional. In addition to surface deformation induced loads, the design of safety features must take into account seismic loaJs and applicable concurrent functional and , accident-induced loads. The design provisions for surface deformation must be based on its postulated occurrence in any direction and azimuth and under any , l part of the nuclear power plant, unless evidence indicates this assumption is ! not appropriate, and must take into account the estimated rate at which the L surface deformation may occur. l 39 i h U
(c) Seismically Induced Floods and Water Waves and Other Design Conditions.- Seismically induced floods and water waves from either locally or distantly generated seismic activity and other design conditions determined pursuant to 5100.23 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 i safety of the public. PART 52 - EARLY SITE PERNITS; STA WARD DESIGN CERTIFICATIONS; AW COMBINED LICENSES FOR NUCLEAR POWER PLANTS
- 7. 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 amended, 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).
- 8. In s52.17, the introductory text of paragraph (a)(1) and paragraph (a)(1)(vi) are revised to read as follows:
552.17 Contents of applications. (a)(1) The application must contain the information required by 5 50.33(a)-(d), the information required by s 50.34 (a)(12) and (b)(10), and to the extent approval of emergency plans is sought under paragraph (b)(2)(ii) of this section, the information required by s 50.33 (g) and (j), and s 50.34 (b)(6)(v)'. The application must also contain a description and safety assessment of the site on which the facility is to be located. 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 s 50.34(a)(1) of this chapter. Site characteristics must comply with Part.100 of this chapter. In addition, the application should describe the following: (vi) The seismic, meteorological, hydrologic, and geologic characteristics of the proposed site; PART 100 - REACTOR SITE CRITERIA
- 9. The authority citation for Part 100 continues to read as follows:
AUTHORITY: Secs. 103, 104, 161, 182, 68 Stat. 936, 937, 948, 953, as amended (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). l
- 10. The table of contents for Part 100 is revised to read as follows:
PART 100 - REACTOR SITE CRITERIA Sec. 40 1
- ~ 100.1 Purpose. 100.2 Scope. ) 100.3 Definitions. 100.4 Communications. I 100.8 .Information collection requirements: OMB approval. l 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 Fa'ctors to be considered when evaluating sites. , 100.21 Non-seismic site criteria. l 100.23 Geologic and seismic siting criteria. APPENDIX A to Part 100 - Seismic and Geologic Siting Criteria for Nuclear Power Plants. ]
- 11. 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 important in assuring that radiological doses from normal 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, and that the site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant. ! (d) This approach incorporates the appropriate standards and criteria for approval of stationary power and testing reactor sites. The Commission 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 important factor in evaluating applications for site approval.
- 12. Section 100.2 is revised to read as follows:
s 100.2 Scope. 41 '
~ c, o .
l i The siting requirements contained in this part apply to applications for ' site approval for the purpose of constructing and operating stationary power l and testing reactors pursuant to the provisions of Parts 50 or 52 of this chapter.
- 13. Section 100.3 is revised to read as follows:
s 100.3 Definitions. ; As used in this part: Combined license means a combined construction permit and operating license with conditions for a nuclear power facility issued puraant to i Subpart C of Part 52 of this chapter. l Early Site Permit means a Commission approval, issued pursuant to : subpart A of Part 52 of this chapter, for a site or sites for one or more ' nuclear power facilities. ExcTusion 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 i protect the public health and safety. Residence within the exclusion area shall normally 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 copulation zone means the area immediately 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 i 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 l such as location, number and size of highways, scope and extent of advance i planning, and actual distribution of residents within the area. ' Population center distance means 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 5s50.21(b) : or 50.22 of this chapter designed to produce electrical or heat energy. : 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 i damping value. The response spectrum is calculated for a specified vibratory motion input at the oscillators' supports. : i 42 !
; O O i
i i The Safe Shutdown Earthauake Ground Motion is the vibratory ground l notion for which certain structures, systems, and components must be designed 4 pursuant to Appendix S to Part 50 of this chapter to remain functional.
! . 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 j
! processes.
- Testina reactor means a testina facility as defined in s50.2 of this l chapter.
) 14. Section 100.4 is added to read as follows: i s100.4 Communications. j Except where otherwise specified in this part, all correspondence, i i reports, applications, and other written communications submitted pursuant to i 10 CFR Part 100 should be addressed to the U.S. Nuclear Regulatory Commission, ' ATTN: Docu^ ment Control Desk, Washington, DC 20555-0001, and copies sent to the appropriate Regional Office and Resident Inspector. Canonications and ! . reports may be delivered in person at the Commission's effices at 2120 L i Street, NW., Washington, DC, or at 11555 Rockville Pike, Rockville, Maryland. }
- 15. Section 100.8 is revised to read as follows: !
s 100.8 Information collection requirements: 0M8 approval. j (a) The Nuclear Regulatory Commission has submitted the information i collection requirements contained in this part to the Office of Management and
- Budget (OMB) for approval as required by the Paperwork Reduction Act of 1995 i (44 U.S.C. 3501 et seq.). OMB has approved the information collection ;
j requirements contained in this part under control number 3150-0093. (b) The approved information collection requirements contained in this
! part appear in s100.23 and Appendix A.
I 1 . 16. A heading for Subpart A is added directly before s100,10 to read as follows:
- Subpart A - Evaluation Factors for Stationary Power Reactor Site i Applications before [ EFFECTIVE DATE OF THIS REGULATION] and for Testing I
Reactors. i
- 17. Subpart B (ss100.20 - 100.23) is added to read as follows:
!; Subpart B - Evaluation Factors for Stationary Power Reactor Site Applications J. on or After [ EFFECTIVE DATE OF THE FINAL RULE). I
- 5100.20 Factors to be considered when evaluating sites.
. 43 ;
i
The Commission 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 must be evaluated to determine whether individual as well as societal risk of potential plant accidents is low, and that site-related characteristics would not prevent the development of a plan to carry out suitable protective actions for members of the public in the event of , emergency. (b) The nature and proximity of marHelated hazards (e.g., airports, dams, transportation routes, military and chemical facilities) must be evaluated to establish site parameters for use in determining whether a plant design can accommodate commonly occurring hazards, and whether the risk of other hazards is very low. (c) Physical characteristics of the site, including seismology, meteorology, geology, and hydrology. (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 maximum probable wind speed and precipitation) must be identified and characterized. (3) Factors important to hydrological radionuclide transport (such as soil, sediment, and rock characteristics, adsorption and retention coefficients, ground water velocity, and distances to the nearest surface body of water) must be obtained from orH;ite measurements. The maximum probable flood along with the potential for seismically induced floods discussed in $100.23 (d)(3) of this part must be estimated using historical data. s 100.21 Non-seismic siting criteria. Applications for site approval for commercial power reactors shall l demonstrate that the proposed site meets the following criteria: i (a) Every site must have an exclusion area and a low population zone, as defined in $100.3; ; (b) The population center distance, as defined in 5100.3, 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 the population center shall be determined upon consideration of population distribution. Political boundaries are not controlling in the application of this guide; (c) Site atmospheric dispersion characteristics must be evaluated and dispersion parameters established such that: (1) Radiological effluent release limits 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 44
(2) Radiological dose consequences of postulated accidents shall meet the criteria set forth in s50.34(a)(1) of this chapter for the type of facility proposed to be located at the site; (d) The physical characteristics of the site, including meteorc, logy, geology, seismology, and hydrology must be evaluated and site parameters established such that potential threats from 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; (f) _ Site characteristics must be such that adequate security plans and measures can be developed; (g) Site characteristics must be such that adequate plans to take protective actions for members of the public in the event of emergency can be developed: (h) Reactor sites should be located away from very densely populated centers. Areas of low population density are, generally, preferred. However, in determining 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, economic, or other factors, which may result in the site being found acceptable'. Geologic and seismic siting factors.
^
s 100.23 This section sets forth the principal geologic and seismic considerations that guide the Commission in its evaluation of the suitability of a proposed site and adequacy of the design bases established in consideration of the geologic and seismic 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 health and safety of the public. Applications to engineering design are contained in Appendix S to Part 50 of this chapter. (a) Applicability. The requirements in paragraphs (c) and (d) of this section apply to applicants for an early site permit or combined license l 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 l
- Examples of these factors include, but are not limited to, such factors as the higher population density site having superior seismic characteristics, better access to skilled labor for construction, better rail and highway access, shorter transmission line requirements, or less environmental impact on undeveloped areas, wetlands or endangered species, etc. Some of these factors are included in, or impact, the other criteria included in this section.
45
, c-,
o . 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) Commencement of construction. The investigations required in paragraph (c) of this section are within the scope of investigations permitted
- by 5 50.10(c)(1) of this chapter.
(c) Geological, seismological, and engineering characteristics. 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 performed 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 must be determined based on the nature of the region surrounding the proposed site. Data on the vibratory ground motica, tectonic surface deformation, nontectonic deformation, earthquake recurrence rates, fault geometry and slip rates, site foundation material, and seirmically induced fToods and water waves must be obtained by reviewing pertinent literature and carrying out field investigations. However, each applicant l shall investigate all geologic and seismic factors (for example, volcanic I activity) that may 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 seismic siting factors. The geologic and seismic siting factors considered for design must include a determination of the Safe , Shutdown Earthquake Ground Motion for the site, the potential for surface l tectonic and nontectonic deformations, the design bases for seismically ' induced floods and water waves, and other design conditions as stated in paragraph (d)(4) of this section. (1) Determination 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 determined considering the results of the investigations required by paragraph (c) of.this section. Uncertainties are inherent in such estimates. These uncertainties must be addressed through an appropriate analysis, such as a probabilistic seismic hazard analysis or suitable sensitivity analyses. Paragraph IV(a)(1) of Appendix S to Part 50 of this chapter defines the minimum Safe Shutdown Earthquake Ground Motion for design. (2) Determination of the potential for surface tectonic and nontectonic deformations. Sufficient geological, seismological, and geophysical data must be provided to clearly establish whether there is a potential for surface deformation. ' (3) Determination 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. 1 (4) Determination of siting factors for other design conditions. Siting factors for other design conditions that must be evaluated include soil and rock stability, liquefaction potential, natural and artificial slope stability, 46 i LL _ _ _ _ - _ __ . _ _ _ _ _
_s - 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 ground motion that may affect the design and operation of the proposed nuclear power plant. l l I Dated at Rockville, Maryland, this day of . For the Nuclear Regulatory Commission. John C. Hoyle, Secretary of the Commission. 47
~ -
4 i i, a 1 1 I l, 4 ATTACHMENT 2 l 1 4 h i 1 l RESOLUTION OF PUBLIC COMMENTS ON THE l i 4
! PROPOSED SEISMIC AND EARTHQUAKE ENGINEERING i
J a CRITERIA FOR NUCLEAR POWER PLANTS 4 1 5 4 4 4 I l i ! 4 i
~ .
O 8 i i 1 1 l RESOLUTION OF PUBLIC COMMENTS i 1 ON THE PROPOSED \ SEISMIC AND EARTHQUAKE ENGINEERING CRITERIA FOR NUCLEAR POWER PLANTS Section 100.23, Geologic and Seismic Siting Factors to 10 CFR Part 100 and Appendix S, Earthquake Engineering Criteria for Nuclear Power Plants to 10 CFR Part 50 October 17, 1994 Publication
COPMENT RESOLUTION
. Factors Section 100.23, Geologic and Seismic Siting to 10 CFR Part 100 and for Nuclear Power Plants Appendix S, Earthquake Engineering Criteriato 10 CFR Part 50 l
B _ACKGROUND ite Criteria Including Seismic and The first proposed revision of the Reactor SPower Plants (10 CFR 20, 1992, (57 FR 47802). Earthquake Engineering Criteria for Nuclear t on October and 100) was published for public ides commen and standard review plan The availability of the draftd regulatory guuidance on meeting Becacse of the the propoj section that were developed to5 provi 1992, e(57gFR 55601). regulations was published on November 2 ,be made in response to substantive nature of the changes to idance documents were withdra the proposed regulations and draft gu f the regulations published for The availability of the replaced with the second proposed 17,1994,(59 FR 52255). revision o i b ry 28,1995, (60 FR 10810). public comment on Octoberdraft guidance documents w t in comments on the October 199 Forty letters (References 1 through 40) con aC iteria for the Seism publication of Proposed Appendix [ Effective 8, " r Date of the Final Siting of Nuclear Power Plants on or AfterCriteria," and/or th Rule)," to 10 CFR Part 100, " Reactor i for Nuclear Site Power Plants," to 10 Appendix S, " Earthquake Engineering Criter ation and Utiliz CFR Part 50, " Domestic Licensing of Produc t ber 17, 1994 (59 FR The Federal Register Notice published on Oc ologic and Seism containing Proposed Section 100.23, "Geo Appendix 8 to 10 10 CFR Part 100 (replacement k Engineering of Proposed Criteria for Nuclear the second Proposed Appendix S, "Earthqua ethe only docum Power Plants," to 10 CFR Part 50 reflect f all comments provid NRC staff evaluation and implementation o 1 to 40. 2
. . l 17, 1994 The resolution of comments contained below relate to the Oct publication.
\
RESOLUTION OF C0194ENTS ON SUPPLEMENTAL INFOR l _ nolicability A la.
"The proposed regulatory action would apply to applicants who a l a construction permit, operating license, preliminar This statement does not l to certification, or combined license ..."41)explicitly indicate whe (Reference the Mined. Geologic Disposal System (MGDS).
lb "The proposed regulatory action would apply to lapplicants who a construction p reit, operating license, This statement does not d apply to preliminar certification, of combined license ..." explicitly indicate whether or not the proposed rev Response. Although Appendix A to 10 CFR Part 100 is titled " Seismic Geologic Siting Criteria for Nuclear Power They Plants," are (1) Partit40, is also referenced in two other parts of the regulation.
" Domestic Licensing of Source Material," Appendix A, " Criterif to the Operation of Uranium Mills and the Disposition of Tailings or l Waste Produced by the Extraction or Concentration of Source from Ores Processed Primarily for Their Source Material Content Section I, Criterion 4(e), and (2) Part 72, " Licensing tive Requireml the Independent Storage of Spent Nuclear Fuel and High-Level R Waste," Paragraphs (a)(2), (b) and (f)(1) of s72.102.
The referenced applicability of Section 100.23 to other than power reactors, if considered appropriate by the NRC, would be a sep rulemaking. That rulemaking would clearly state the applicability of In addition, NUREG-1451 or other facility. Section 100.23 to a MRS will remain the NRC staff technical position on k seismic siting iss pertaining to a MGDS until it is superseded through a rulem revision of NUREG-1451, or other appropriate mechanism. 3
_y e < l Section VfB)(3) " Uncertainties and Probabilistic Methods"
- 1. It is. stated that "Because so little is known about earthquake
. phenomena..." Use of the expression "so little is kno W creates a 1 t
false impression of the current state of knowledge about earthquake phenomena. Although our understanding of earthquake phenomena remains uncertain,-quantum advances in knowledge have been made during the past 25 years. With these very significant advances, geoscientists now have much more confidence than previously in expressions of uncertainty ! regarding interpretations of inputs to a probabilistic seismic hazard ! anabses; and these can be fully accounted for in the uncertainty in the seismic hazard results. The language of the regulation should reflect 1 these very positive developments. (Reference 41) I Resnonse: The statement will be revised to put less emphasis on the negative as follows: "Because of uncertainties about earthquake i pheniana (especially in the eastern United States), there have often been differences of opinion and differing....."
- 2. The key elements of the NRC's proposed balanced approach are listed.
The wording of the fourth element should be revised to indicate that the 9eoscience investigations refer to site-specific data, or new regional data, or a combination of the two. (Reference 41) Response: It refers to both regional and site investigations. The l element will be revised to: " Determine if information from the regional and site geoscience investigations....." Section VfB)(5). "Value of the Operatina Basis Earthauake Ground Motion (OBE) and Reauired OBE Analysis." Does not support the NRC staff's position to not require explicit design analysis for the Operating Basis- Earthquake Ground Motion (OBE). The staff's position is not sound, not technically justified, and not appropriate for the design of Section III pressure-retaining components. It is not possible to inspect to verify that cyclic fatigue effects for the OBE are insignificant. There is no technical basis to state that OBE should not control the design of safety systems. It is not technically . justified to assume that Section III components will remain within applicable stress limits at one-third of the SSE. Equipment necessary for continued operation, but not required for safe shutdown, is not required to be designed for OBE nor SSE. The following specific comments [1 through 7] pertain to the , supplemental information to the proposed regulations, item V(B)(5),
"Value of the Operating Basis Earthquake Ground Motion (OBE) and Required OBE Analysis." Comments are limited to the design of pressure-retaining components to the ASME Boiler and Pressure Vessel Section III rules. (Reference 42) 4
~
- 1. Regarding the soundness of SSE only design:
"For instance, the NRC staff, SECY-79-300, suggested that design for a single limiting event and inspection and evaluation for earthquakes in excess of some specified limit may be the most sound regulatory approach."
This is not a sound regulatory approach if it is not feasible to inspect for cyclic damage to all the pressure-retaining components. It is not feasible to inspect. Many components are not accessible. Even if accessible, the components may be covered with insulation. Even if there is not insulation or the insulation is removed, it is not feasible to inspect to determine the amount of the fatigue life used by the OBE cyclic loads. It is not feasible to inspect for crack initiation on the inside of the component in all critical areas. Even if it were feasible to inspect for cracks, it is possible to have an unacceptable amount of fatigue life used by the OBE without crack initiation. Visually inspecting for permanent deformation, or leakage, or failed component supports is certainly not adequate to determine cyclic damage. Response. SECY-79-300, " Identification of Issues Pertaining to Seismic and Geologic Siting Regulation, Policy, and Practice for Nuclear Power - Plants," informed the Commission of the status of the staff's reassessment of Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part 100, " Reactor Site Criteria." The cited statement appeared in an enclosure (Enclosure B, Section 2.4) discussing issues arising from engineering requirements in Appendix A, procedures for providing an interface of these requirements with geologic and seismic input, and with matters involving scientific and engineering conservatism. In a related area (Enclosure A, Section 2.4), the NRC staff informed the Commission about problems in applying the Appendix A requirement that the plant must be shut down and inspected if ground motion in excess of that corresponding to the OBE occurs because there is no definitive shutdown guidance or inspection criteria. l The proposed regulations is similar to the statement in SECY-79-300 in that it allowed plants to be designed for a single limiting event (the SSE) and inspected and evaluated for earthquake in excess of some specified limit (the OBE) when and if it occurred. Also, the proposed regulation allowed for the plant to be designed at both the SSE and OBE level s. Earlier concerns expressed in SECY-79-300 regarding OBE exceedance and shutdown / restart guidelines have been resolved. A criterion to determine OBE exceedance is described in Regulatory Guide 5
_ - . , . . . , . . . . . . . ~ ..~.-~_...m .-_w ,. g e 4 1.166, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Postearthquake Actions," (Draft was DG-1034). Postearthquake inspection and evaluation guidance is described in Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by an Seismic Event,"
-(Draft was DG-1035). The guidance is not limited to visual inspections, it includes inspections, tests, and analyses including fatigue analysis.
- 2. Regarding OBE controlling design:
"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 OBE should not control the design of safety systems." There is no technical basis for stating that the OBE should not control the design of safety systems. Based on my knowledge of current plant designs, I can state that if there are five OBE's of the magnitude of one-half the SSE expected to occur in the life of the plant, then OBE will control the design of the piping systems. And in this case, OBE should control the design. The cyclic effects of the repeated earthquakes have to be considered in the design of the component to ensure pressure boundary integrity throughout the life, especially if the SSE can occur .after the lower level earthquakes. 1 The appropriate action is to define the magnitude of the OBE that is , expected to occur, and to require the component manufacturer to design for the OBE. It appears that NRC is assuming the liability for the proper design of a pressure-retaining component for a lower level earthquake. It should be the N certificate holder's responsibility to provide a component that is structurally and functionally adequate for both the OBE and the SSE. Resnonse. The NRC staff agrees that the cyclic effects of repeated earthquakes have to be considered in the design of the components to ensure pressure boundary integrity. The NRC staff has identified actions necessary for the design of structures, systems, and components l when the OBE design requirement is eliminated (these actions include fatigue analysis). A discussion pertaining to these actions (provided in SECY-93-087, Issue I.M), is included within supplemental information l item V(B)(5) of the proposed regulation. The guidelines in SECY-93-087 provide a level of fatigue design for the piping equivalent to that currently provided in the Standard Review Plan Section 3.9.2. l f- LAlso, The NRC staff has concluded that design requirements based on an L estimated OBE magnitude at the plant site and the number of events 6
m ,. expected during the plant life will lead to low design values that will not control the design thus resulting in unnecessary analyses.
- 3. Regarding explicit response or design analyses:
"The proposed regulation would allow the value of the OBE to be set at (1) one-third or less of the SSE, where OBE requirements are satisfied without an explicit response or design analysis.. "
The OBE requirements are - "... componer.ts .... 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." It is not~ technically justified to assume that Section III components will-remain 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 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. For Level B, the component must be designed to withstand the cyclic effects of the earthquake load and all other cyclic Level A and B loads without damage requiring repair. In order for the assumption to be valid -- that at one-third SSE, the Level B criteria is satisfied for a component designed for the SSE -- the cyclic fatigue damage from the OBE must be insignificant. It is highly improbable that the fatigue damage from the OBE will be insignificant unless the component is designed for the OBE. Response. The following is extracted from SECY-93-087, " Policy, l: Technical and Licensing Issues Pertaining to Evolutionary and Advanced l-Light-Water Reactor (ALWR) Designs," Issue I.M, " Elimination of Operating-Basis Earthquake."
"A designer of piping systems considers the effects of.
l primary and secondary stresses and evaluates fatigue caused by repeated cycles of loading. Primary stresses are induced j by the inertial effects of vibratory motion. The relative motion of anchor points induces secondary stresses. The ; repeating seismic stress cycles induce cyclic effects I (fatigue). After reviewing these aspects, the staff concludes that, for primary stresses, if the OBE is established at one-third the 7
. e
. o SSE, the SSE, load combinations control the piping design when the earthquake contribution dominates the load combination. Therefore, the staff 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 OPE loadings Accordingly, to account for earthquake cycles in the fatigue analysis of piping systems, the staff proposes to develop guidelines for selecting a number of SSE cycles at a fraction of the peak amplitude of the SSE. These guidelines will provide a level of fatigue design for the piping equivalent to that currently provided in the standard review plan (SRP) (NUREG-0800).*
Positions pertaining-to the elimination ~of the Operating Basis Earthquake were proposed in SECY-93-N7. Commission approval is documented 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. Regarding the OBE and PRA insights:
"There is high confidence that, at this ground-motion 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 margins-type assessment of seismic events."
8
. . - . _ _ _ . _ _ _ _ . _ _ . . _ _ _ _ . ~ . _ _ . . _ _ _ . _ _ _-____
3 This technical position is not valid for Section III' pressure-retaining components. . As stated under comment 3, cyclic effects are not considered for the SSE. There is no possible way to predetermine that
.the cyclic effects at one-third SSE are insignificant without evaluating ,
_ specific configurations. .To say that PRA insights from a margins-type l assessment will ensure that Level B design limits will be satisfied at one-third SSE is completely wrong. Response. See response to comment 3. 1
- 5. Regarding NRC proposed criteria:
"Also, the NRC staff has evaluated the effect on. safety of i eliminating the OBE from the design load combinations for selected i structures, systems, and components and has developed proposed l
, criteria for an analysis using only the SSE." J l
The proposed criteria referred to is the proof that "SSE only" is not a , prudent regulatory approach. In order to ensure that the OBE l requirements are satisfied at one-third SSE, the NRC staff is requiring a fatigue evaluation for two SSE's for the ABWR. This may be more restrictive than designing for five OBE's at one-third SSE. Consider what has happened. The NRC staff realized that it is not sufficient for i Section III components to be designed only for the SSE. They are requiring an explicit fatigue analysis _ so that the OBE requirements will be satisfied. The bottom line is that the NRC' staff, in implementing
"$SE only," have required an explicit for an ' equivalent OBE loading. A better approach would be to design for the OBE.
Response. The proposed criteria is a prudent regulatory approach. On the basis of analysis, tests, and engineering judgement, the NRC staff has determined the design produced using SSE load combinations, in general, envelop the load combinations produced using the OBE. For specific situations such as~ piping, where eliminating the OBE will j directly affect the current methods used to evaluate the adequacy of cyclic and secondary stress effects in the piping design procedures have been developed (see response to comment 3).
- 6. Regarding required plant shutdown:
" Prior to resuming operations, the licensee will be required to demonstrate to the Commission that no functional damage has occurred to those features necessary for continued operation without undue risk to the health and safety of the public."
If the applicant does not do an analysis and design for one-third SSE, the applicant is required to shutdown and inspect if the one-third SSE occurs. Obviously, the assumption is that the applicant can inspect to determine if there is damage to the Section III components. It is not 9
n
. .4 i
i possible'to inspect to determine if there is cyclic damage to the- l Section III. pressure-retaining components. The damage that has to be assessed is the effect of the cyclic loads on the life of the component. You are not inspecting for permanent deformations, leaks, or bent or failed supports. If these conditions occur at one-third SSE, then the plant seismic design is obviously deficient. You need to determine that the cyclic effects are not significant. This is impossible to determine by ' inspection. The question that has to be answered it whether the fatigue usage factor from the.0BE is acceptable. The acceptability of the fatigue usage factor for a specific component is dependant on the i severity of all the other cyclic loads on the component. The cyclic l effects from the'0BE for a component with high fatigue damage from ' service conditions, a pressurizer surge line or a nozzle subject to flow ' stratification effects for example, would have to be insignificant. The fatigue " damage" from the OBE cannot be determined by inspection.
~
Analysis is the only method to verify that the OBE cyclic effects are within acceptable limits. The only reasonable approach is to perform ; the '0BE fatigue analyses as part of the component design process. Response. Postearthquake inspection and evaluation guidance is described in Draft Regulatory Guide DG-1035, " Restart of a Nuclear Power Plant Shut Down by an Seismic Event." The guidance is not limited to , visual inspections, it includes inspections, tests, and analyses including fatigue analysis.
- 7. Regarding equipment seismic design: -{
"The Operating Basis Earthquake Ground Motion (OBE) 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 Safe Shutdown Earthquake Ground Motion (SSE) is the vibratory ,
ground motion for which certain structures, systems, and components must be designed to remain functional." [Three types of : equipment are described.] There is one major flaw in tha "SSE only" design approach. The ' equipment 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 "necessary 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 OBE effects. I am disappointed that a proposed rule would be published with flaws in . the technical logic. Perhaps the approach of designing for the SSE only ' is adequate for building structures designed to AISC rules, but this approach is certainly not adequate for Section III pressure-retaining i 10
components. There appears to be a lack of understan' ding of the Section III design requirements and the significance of seismic loads. To assume that the component stresses will be within the Section III Level B code requirements at 1/3 the SSE if the component is designed for the , SSE is not valid. To assume that an applicant can properly inspect the ; safety related components after an OBE earthquake to determine that the ' ability of the components to function for the remaining life has not ; been impaired is unreasonable. The potential problem is detrimental ! impact on the fatigue life from the cyclic OBE loading. There is no feasible way to inspect for detrimental impact on fatigue life. It is not prudent to design only for SSE, and to assume that there will be no cyclic damage from the OBE. I see no reason to compromise the seismic design of the plant. It is inappropriate to assume that design . for OBE is not required without even knowing the component configuration. We d'o have a problem in the industry with the present requirements. ' Requiring design for five OBE events at h SSE is unrealistic for most (all?) sites and requires an excessive and unnecessary number of seismic , supports. The solution is to properly define the OBE magnitude and the : I 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 t you cannot assume, before you have the seismicity defined and before you have a component design, that OBE will not govern the design. , The problem with not designing for OBE can be simply stated. The pressure-retaining component may be designed to the fatigue limit for other Level A and B loads (for example, thermal transients). In this situation, OBE stresses above the endurance limit reduce the operational i 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 j 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 l of seismic supports. 1 The rule refers to "new information and research results." The newest information and research results is the Northridge earthquake and the Kobe earthquake. In the Northridge earthquake, steel building members : cracked and this behavior was unexpected. In the Kobe earthquake, a seismically designed elevated highway toppled over, and this behavior l was unexpected. What I have learned from these events and earlier i earthquakes, is that our understanding of seismic response is limited. l Conventional wisdom is that ductile steel piping systems will not fail
- in a single earthquake event. But in a recent NRC/EPRI program on l dynamic reliability, undegraded piping components failed in a single earthquake event. The loadings were extreme in most cases, but the failure in a single event was not expected, t
- The intent of the rule making, to uncouple the OBE and the SSE, is a l necessary change in the seismic requirements.
i 11
- e i
Response. It is not possible that some equipment necessary for f
- continued n ft operation will not be designed for SSE or OBE effects.
- j. General Design Criterion 2, " Design Bases for Protection Against Natural Phenomena'," of Appendix A, " General Design Criteria for Nuclear Power i Plants," to 10 CFR Part 50 requires that nuclear power plant structures, I 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 1 implement General Design Criterion 2 insofar as it requires structures, systems, and components important to safety to withstand the effects of earthquakes. Regulatory Guide 1.29, " Seismic Design Classification," describes a method acceptable to the NRC staff for identifying and l 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 the Supplemental Information, Section VII, Future Regulatory Actions, regulatory guides related to these components will be revised to provide alternative design requirements. See response to comments 3 and 5 for discussions on stress limits and fatigue. RESOLUTION OF COMMENTS ON SECTION 100.23 (a) Anolicability.
- 1. The language relevant to an applicant under Part 50 appears to be intended to avoid "backfitting" the new criteria in lieu of that used to obtain the construction permit originally. Unfortunately, the words sha77 comply unnecessarily imposes retention of the original Appendix A criteria on such applicants. Although unlikely, an applicant already holding a construction permit may elect to apply the new methodology and criteria. Replace "shall comply" with "may elect to demonstrate i compliance with the seismic and geologic siting criteria in Subpart A or !
B to Part 100 of this Chapter." (Reference 43) l 1 Response. The NRC will address this request on a case-by-case basis rather than through a generic change to the regulations. This situation 12
l l l
- t. 1
! pertains to a limited number of facilities in various stages of l l construction. Some of the issues that must be addressed by the l applicant and NRC during the operating license review include I l differences between the design bases derived from the current and i amended regulations (Appendix A to Part 100 and Section 100.23, respectively), and earthquake engineering criteria such as, OBE design l requirements and OBE shutdown requirements. l (d)(1) Determination of the Safe Shutdown Earthauake Ground Motion.
- 1. Determination of the SSE is based upon an evaluation that includes investigation of geological and seismological information and the results of a probabilistic seismic hazard analysis. Addressing uncertainties is an inherent part of the process.
Based upon prior licensing decisions and scientific evaluations (Systematic Evaluation Program, Appendix A evaluations, LLNL, and EPRI) it seems reasonable to only perform detailed confirmatory site investigations (Regulatory Guide 1.132) at existing sites. Standardized 0.3g advanced plant designs are sufficiently robust to bound the seismic l design attributes of all nuclear power plants at current sites. Inclusion of these simplified requirements for existing sites represents a significant step toward predictable and cost-effective licensing. Revise to read (substitution in italics): " Determination of.the Safe Shutdown Earthquake Ground Motion. The Safe Shutdown Earthquake Ground Motion for the site is characterized by both horizontal spectra and vertical free-field ground motion response spectra at the free ground surface. The Safe Shutdown Earthquake Ground Motion for the site is based upon the investigations required by paragraph (c) of this section and the results of a probabilistic seismic hazani analysis. Selsnological and geological uncertainties are inherent in these determinations and are captured by the probabilistic analysis. Suitable sensitivity analyses'nay also be used to evaluate uncertainties. Paragraph IV (a)(1) of Appendix S to Part 50 of this Chapter defines the minimum Safe Shutdown Earthquake Ground Motion for design. Based upon prior scientific findings and ifcensing decisions at existing nuclear power plant 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 foundation evaluations." (Reference 43) Egigonig. (1) Determination of the Safe Shutdown Earth Ground Motion. Your recommended rewording is another way of saying the same thing, but places less emphasis on site-specific investigations relative to the PSHA than the current wording. We regard the current wording as better j reflecting the proper priorities. Site specific investigations (regional' and site geological, seismological, geophysical, and 13
o , 4 l geotechnical) are of prime importance in deriving the bases for the SSE. It must not be forgotten that if all of the data that is needed about a site to determine the SSE could be obtained through site-specific j investigations, a PSHA would not be necessary. However, because of ! 1 uncertainties, at the present time, more reliance must be placed on ' PSHA's than may be necessary in the future when more information is available. Paragraph IV(a)(1) of Appendix S to Part 50. Investigations at most of the existing sites will imore than likely be confirmatory if the initial investigations were thorough, and there has not been too much lag time since the initial investigations were accomplished and the results reviewed by the NRC. However, in many cases it may be necessary to carry out more extensive investigations than are usually considered as
" confirmatory" investigations because: (1) the state-of-the-science is rapidly changing as new information is derived from every earthquake that occurs, and from ongoing research; (2) applicants may elect not to use the standard design plant and justify an SSE different than 0.3g; and (3) it will often be necessary, even for standard design sites, to determine a site-specific SSE as the design basis for other, non-
^
standard design, safety-related structures, systems or components such as dams, reservoirs, intake and discharge facilities, etc. The current wording in the proposed regulation most accurately represents the NRC staff's position on this issue.
- 2. Proposes that at existing eastern U.S. sites (rock or soil), or at eastern U.S. rock sites not located in areas of high seismicity (for example, Charleston, South Carolina, New Madrid, Missouri, Attica, New York) a 0.3g standardized ALWR design is acceptable and only evaluations of foundation conditions at the site are required (Regulatory Guide 1.132), but not geologic / geophysical seismological investigations. For other sites a DG-1032 review is required.
Proposes that-10 CFR Part 100 Section 100.23(d)(1) be modified to reflect this consideration as follows:
" Determination 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 based upon the investigations required by paragraph (c) of this 14
_ _. . _ . - _ _ _ _ . _ _ _ _ _. _ - _ _ . _ _ _ _~
.- . i section and the results of a probabilistic seismic h'azard analysis.
- Seismological and geologic uncertainties are inherent in these determinations and are captured by the probabilistic analysis. Suitable
. sensitivity analyses may also be used to evaluate uncertainties.
l Paragraph IV(a) (1) of Appendix S to Part 50 of this Chapter defines the minimum Safe Shutdown Earthquake Ground Motion for design. Based upon prior scientific findings and licensing decisions at existing nuclear l power plant sites east of the Rocky Mountain Front (east of approximately 105 west longitude) a 0.3g Standardized design level is l acceptable at these sites given confirmatory foundation evaluations. l For rock sites not in areas of known seismic activity including but not L limited to the regions around New Madrid, MO, Charleston, SC, and i Attica, New York, a 0.3g Standardized design level is acceptable given ;
- confirmatory foundation evaluations at the site." (Reference 44) !
l l Response. Although some of the suggested wording may improve the l readability of the text, the staff does not agree with the basic philosophy of the recommended modification for the following reasons:
- 1. The suggested modification brings back a prescriptive element which we have tried to eliminate in revising the siting document.
It is more appropriate to include such a modification in Regulatory Guide 1.165 (Draft was DG-1032). The staff's position regarding the application of the 0.3g ALWR design is addressed in the main body of the draft guide, and in Appendix D. l 2. A standard design of 0.3g does not preclude the need to conduct a thorough regional and site' area investigation. The standard plant is designed for 0.3g, but other safety related components aren't part of the standard design plan. Such components include emergency cooling' ponds and associated dams levees, spillways, r etc., and they will have to be designed to the appropriate level based on regional and site geological, seismological, geophysical, ! and geotechnical investigations. 3.- The level of investigations for a standard design plant or any l additional unit sited on a previously validated site depends on when that site was previously validated, the complexity of the geology and seismology of the region and site, the advent of new information or hypotheses about regional tectonics, and the kinds of methods used and the thoroughness applied in using those 15
4 9
+
methods in the original investigations and analyses. The investigations can range anywhere between a literature review to a-
,very extensive investigation program.
- 4. The discovery 'of the Meers Fault and the paleoseismic evidence for a large prehistoric earthquake-in the Wabash Valley are examples in the central and eastern U.S. of the occurrences of events of l great' significance to the seismic hazard to those regions that were unknown until regional investigations were performed. Thus,
)
we expect that evidence for siellar,- currently unknown tectonic ; structures or events is present in the CEUS. Based on the above factors, the level of investigations could vary considerably, therefore, it would be inappropriate to make the modifications recommended. RESOLUTION OF COMENTS ON APPENDIX S TO PART 50 General Information
- 1. ' Mandate the retrofit of existing nuclear power plants in extremely i
-active seismic zones with the most recent ASCE seismic design and i engineering criteria. The requirements should be phased in a manner to take effect at individual reactors at the time of relicensing to ease the financial impact on the licensees. (Reference 45)
Resoonse. This regulation is applicable to applicants for a design certification, combined license,-construction permit or operating
-license on or-after the effective date of the final rule. Because the requested change pertains to existing. (operating) nuclear power plants it is beyond the scope of this rulemaking. The regulations pertaining .
to relicensing are contained in 10 CFR Part 54, " Requirements for Renewal of Operating Licenses for Nuclear Power Plants." Further, If the.NRC staff were to change the licensing bases for operating plants the burden would be on the' staff to ensure that the backfit requirements stated in Section 50.109, "Backfitting," to 10 CFR Part 50, " Domestic
. Licensing of Production and Utilization Facilities," are met.
16
- 2. - There are several phrases that are used in the regulation that should be modified to make-the regulation more stable from a licensing point of view. The following phrases and others that are similar in nature
-should be modified: . (Reference 46) 2a. "
... certain structures, systems, and components ..." should read:
... certain structures, systems, and components as identified in Regulatory Guides XXX ..." By referencing the regulatory guides,
. the vagueness of the statement is eliminated from the rule and the description of the structures, ::ysterns and components can be changed, if necessary, via changes to the regulatory guides."
Response. Regulatory guides are issued to describe and make available to the public such information as methods acceptable to the NRC staff for implementing specific parts of the Commission's regulations, techniques'used by staff in evaluating specific problems or postulated accidents, and guidance to applicants. The Introduction section of the I guide cites the applicable regulations _ pertaining to the guidance. Regulatory guides are not cited in regulations. The regulation was not i changed. ) i 2b. "
... without loss of capability to perform their safety functions" should read: " ... without loss of capability to perform their intended functions." The components perform a function and not a f " safety" function -- components may be part of a safsty system or l a non-safety system. There are other sentences which have a similar phraseology -- for example, item c below. These sentences should be similarly modified.
Response. The term " safety function" is synonymous with terminology codified in other regulations; for example, General Design Criterion 2,
" Design Bases for Protection Against Natural Phenomena," of Appendix A,
" General Design Criteria for Nuclear Power Plants," to 10 CFR Part 50.
The regulation was not changed. 2c. "The required safety functions of structures, systems, and components must be assured ..." should read: "The required functions of structures, systems, and components must be assured per the guidelines provided in Regulatory Guide XXX ..." The change shows that the regulatory guide contains guidance as to how a future license applicant can provide " assurance." , Response. See response to comments 2(a) and 2(b). The regulation was not changed. 17
9 1
~ j l
Definitions i
- 1. The parenthetical phrase in the definition of response spectrum should
.be changed to (acceleration, velocity, and displacement) (not "or" displacement]. Displacement is also involved in a response spectrum.
(Reference 41) Response. There are situations where it is only necessary for the ' response spectrum plot to show one of the three parameters depicted; for example, a plot of accelerations and frequencies. The definition was not changed. I Safe Shutdown Earthauake Ground MoiM g
- 1. Incorporate the seismic design 7.nd engineering criteria of ASCE Standard
'4, " Seismic Analysis of Safety-delated Nuclear Structures and Commentary on Standard for Seismic Analysis of Safety-Related Nuclear Structures,"
into Part 100 to strengthen the basis for the requirements. (Reference 45) Response. The supplemental information to the proposed regulations, l item VB(2), " Remove Detailed Guidance from the Regulation," cites that i the current regulation.(Appendix A to 10 CFR Part 100) is too detailed, ) containing both requirements and guidance to satisfy the requirements. It further notes that having detailed assessments cast in a regulation ) has caused difficulty for applicants and the NRC staff in terms of. ! inhibiting the use of needed latitude in judgement. Also, it has ) inhibited flexibility in applying basic principals to new situations and the use of evolving methods of analysis (for instance, probabilistic) in th3 licensing process. Therefore, the Commission has determined that new regulations will be more streamlined containing only basic requirements with guidance being provided in regulatory guides and, to some extent, in standard review plan sections. Therefore, it is common 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 in regulatory guides and standard review plan sections. ASCfStandard4iscitedinthe1989revisionof Standard Review Plan Sections 3.7.1, 3.7.2, and 3.7.3. Operatina Basis Farthaualgtfgound Motion
- 1. Support <, the NRC staff's position to not require explicit design
. analysis for the Operating Basis Earthquake Ground Motion (0BE) if its 18 l
1
peak acceleration is less than one-third of the Safe' Shutdown Earthquake Ground Motion (SSE). The OBE for ABB-CE's System 80+" is less than l one-third of the SSE. The supporting analysis has already been reviewed <
.and approved by the NRC staff in NUREG-1462, " Final Safety Evaluation '
i Report Related to the Certification of the System 80+ Design." (Reference 47) Surface Deformation
- 1. There is no definite indication of the type of deformation that must be 4 considered. A clear distinction should be made between tectonic and on-tectonic deformation; and the design actions appropriate for both provided. (Reference 41) t Response. The definition of surface deformation in Appendix S to 10 CFR Part'50 addresses tectonic surface deformation as a subset of surface deformation. Therefore, it is not necessary for the discussion in the regulation (Paragraph IV(b)) to distinguish between surface tectonic and nontectonic deformations. In addition, Section 100.23(d), " Geologic' and Seismic Siting Factors," to 10 CFR Part 100 requires, in part, that the geologic and seismic siting factors considered for de' sign include the potential for surface tectonic and nontectonic deformations.
l l With regard to including a discussion on design actions appropriate for i both surface tectonic and nontectonic deformations, the Commission has determined that new regulations will be more streamlined containing only { basic requirements; guidance will be provided in regulatory guides and, to some extent, in standard review plan sections as appropriate. ; Therefore, design actions will not be provided in the regulation. The response to comment C1 contains additional discussion on the removal of f detailed guidance from the regulation.
- 2. The required consideration of aftershocks is confusing and not needed.
It has been recognized from early.in the NRC's implementation of seismic design requirements that design for the SSE is more than adequate to account.for any vibratory ground motion due to aftershocks. , Alternatively, clarifying language should be added indicating aftershocks are fully considered in SSE design. (Reference 41) , Response. The reference to aftershocks will be deleted. One of the changes to the Appendix A to Part 100, Safe Shutdown Earthquake requirements was the deletion of the phrase " including aftershocks." ' i 19 -
.. - - ~ , ,
The recommended change will make the aftershock requirements in Paragraphs IV(b), " Surface Deformation, and IV(a)(1), " Safe Shutdown
. Earthquake Ground Motion," of Appendix S to 10 CFR Part 50 consistent.
- 3. When surface deformation is identified as a hazard at a site, the :
determination of appropriate design parameters will specifically include a determination of its spatial characteristics. The requirement to postulate the occurrence of the load in any direction and azimuth and
+
under any part of the nuclear plant is inappropriate, and should be removed. (Reference 41) Response. The regulation specifically states if and how spatial i characteristics for surface deformation must be considered in design. The same requirements are contained in Paragraph VI(b)(3) of Appendix A j to Part 100 (effective December 1973). A technical justification j stating why it is inappropriate to require the postulated occurrence of ;
'the load in any directien and azimuth and under any part of the nuclear !
plant was not provided. Tile regulation was not changed. l l 4 20 J
,w-
. s REFERENCES
- 1. Republic of China Atomic Energy Council, Tsing-Tung Huang, February 18, 1993
- 2. Korea Electric Power Corporation, Chung, Bo Hun, December 22, 1992
- 3. Corps of Engineers, Ellis L. Krinitzsky, March 1,1993 I l
- 4. Association of Engineering Geologists, Jeffery R. Keaton, March 5,1993
- 5. W. Scott Dunbar, March 3, 1993 l
- 6. Ohio Department of Natural Resources, Michael C. Hansen, February 16, '
1993
- 7. North Dakota Geological Survey, John P. Bluemle, January 19, 1993
- 8. Federation of Electric Power Companies, Ryo Ikegame, March 15, 1993
- 9. Electricite de France, Remy Carle, March 10, 1993
- 10. New York Power Authority, Rfalph E. Beedle, March 18, 1993
- 11. Scottish Nuclear Limited, R.J. Killick, March 22, 1993
- 12. G.C. Slagis Associates, Gerry C. Slagis, March 22, 1993
- 13. Ohio Citizens for Responsible Energy, Inc. (0CRE), Susan L. Hiatt, March 22, 1993
- 14. Yankee Atomic Electric Company, D.W. Edwards, March 23, 1993
- 15. California Department of Conservation, James F. Davis, March 23, 1993
- 16. Georgia Power Company, J.T. Beckman, Jr., March 24, 1993
- 17. Southern Nuclear Operating Company, J.D. Woodard, March 24, 1993
- 18. Virginia Power, William L. Stewart, March 22, 1993
- 19. ENEA-DISP, Giovanni Naschi, March 24, 1993
- 20. Nuclear Management and Resources Council, (now Nuclear Energy Institute)
William H. Rasin, March 24, 1993
- 21. Department of Energy, Dwight E. Shelor, March 24, 1993
- 22. Westinghouse Electric Corporation Energy System, N.J. Liparulo, March 24, 1993
- 23. Niagara Mohawk.Powar Corporation, C.D. Terry, March 24, 1993
- 24. GE Nuclear Energy, P.W. Marriott, March 23, 1994 21
f V f)
- 25. Gulf States Utilities Company, J.E. Booker, March 24, 1993
- 26. Nuclear Electric, B. Edmondson, March 31, 1993
- 27. Florida Power & Light Company, W.H. Bohlke, March 24, 1993
- 28. Nuclear Electric, B. Edmondson, March 31, 1993
- 29. Ministere de L'Industrie etc, Michel Laverie & Walter Hohlefelder, March 31, 1993
- j. 30. Deleware Geological Survey, Thomas E. Picket, March 10, 1993
- 31. Tennessee Valley Authority, Mark J. Burzynski, April 22, 1993
- 32. Florida Power Corporation, Rolf C. Widell, April 23, 1993
- 33. Depa'rtment of Energy, Jeffrey Kimball, May 17, 1993 l 34. National Atomic Energy Agency, Djali Ahimsa, May 26, 1993
- 35. Norman R. Tilford, May 27, 1993 I
- 36. International Siting Group, William 0. Doub, June 1,1993
- 37. U.S. Geological Survey, Dallas L. Peck, June 2,1993 l 38. American Nuclear Society, Nuclear Power Plant Standards Committee,
! Walter H..D'Ardenne, June 24, 1993
- 39. Sargent & Lundy Engineers, B.A. Elder, March 23, 1993
- 40. EQE International, James J. Johnson, July 12, 1993
- 41. U.S. Department of Energy, Ronald A. Milner, May 11, 1995
- 42. G.C. Slagis Associates, Gerry C. Slagis, February 13, 1995
- 43. Nuclear Energy Institute, William H. Rasin, May 12, 1995
- 44. Yankee Atomic Electric Corporation, Stephen P. Schultz, May 19, 1995 l
l 45. American Society of Civil Engineers, Washington Office, Stafford E. l Thornton, February 14, 1995 l 46. Westinghouse Electric Corporation, N.J. Liparulo, June 2,1995
- 47. ABB Combustion Engineering Nuclear Systems, C.B. Brinkman, May 8, 1995 22
i i, i i i i ATTACHMENT 3 ACRS LETTER I l l l l 1 t i
/ %o UNITED STATES f NUCLEAR REGULATORY COMMISSION i 4
5 I ADVISORY COMMITTEE ON REACTOR SAFEGUARDS ' O, [ WAS$44NGTON, D. C. 20806 y j
%, * . . .
- f April 22, 1995 i
The Honorable Shirley Ann Jackson Chairman 4 U.S. Nuclear Regulatory Commission } a Washington, D.C. 20555-0001
Dear Chairman Jackson:
SUBJECT:
PROPOSED REVISIONS TO 10 CFR PARTS 50 AND 100 AND PROPOSED REGULATORY GUIDES RELATING TO REACTOR SITE . j CRITERIA I During the 430th meeting of the Advisory Committee on Reactor Safeguards, April 11-13, 1996, we reviewed the proposed revisions j ! i to reactor siting regulations and associated Regulatory Guides and . Standard Review Plan sections. Our Subcommittee on Extreme l External Phenomena reviewed this matter during a meeting on April ! 4 3, 1996. During this review, we had the benefit of discussions-
- with representatives of the NRC staff, Westinghouse Electric j Corporation, and the Nuclear Energy Institute. We also had the i
benefit of the document referenced. f 1 The staff has proposed final revisions to 10 CFR Parts 50 and 100 ' ! and a new Appendix S to Part 50 that deal with both seismic and i source term issues for future plants and sites. Many of the implementation details will be found in new Regulatory Guides and
- in Standard Review Plan sections. The existing requirements.of 10 i CFR Part 100 and its Appendix A will remain in effect for operating l plants.
- We recommend that the proposed final rule dealing with the seismic
] aspects be issued. i j The proposed final rule requires that any individual, located at 4 any point on the exclusion area boundary for any two-hour period ! following the postulated release of the fission products, not
- receive a radiation dose in excess of 25 ren total effective dose equivalent (TEDE). Similarly, an individual located at the outer boundary of the low population zone '(LPZ), who is exposed to the
; radioactive cloud resulting from the release of the postulated fission products (during the entire period of its passage), not
- j. receive a dose in excess of 25 ram TEDE. Consistency within the
! body of NRC regulations is most desirable. We recommend that ! careful definitions of the TEDE limits that are mindful of organ i dose weighting factors found in 10 CFR Part 20 be included in the final rule.
I
-.. - . - - - _ - . - . ~ . - . . ~~~
L , , I l Radiological doses are to be evaluated over a two-hour period. The proposed final rule states that the evaluation should be over the two-hour period of maximum dose. The Office of Nuclear Regulatory Research (RES) has a differing view and recommends that the proposed final rule be modified from any two-hour period after release of fission products (referred to as the " worst" two hours) to a period of two hours commencing with fuel failure (referred to as the "first" two hours). RES beljeves that the use of the worst two-hour period in the dose calculation is not justified by risk considerations and could lead to increased costs for future licensees with no commensurate gain in safety. The staf.f supporting the proposed rule states that (1) the proposed licensing framework would provide a relaxation of engineered safety feature (ESF) performance requirements commensurate with updated > source term and radiological insights, (2) the regulatory requirements for determination of in-containment radioactive
" material during the two-hour dose evaluation period would be ;
consistent and capable of handling designs substantially different from those analyzed in NUREG-1465, " Accident Source Terms for Light-Water Nuclear Power Plants," (3) the analysis would be easy to perform and reprodacible with confidence, and (4) the technical bases and analytical methods would be defensible. While the revised dose evaluation in 10 CFR 50.34 is intended for future 3 plants, the staff is concerned that a current licensee might seek l to use . it to remove or disable existing fission product cleanup systems. This could markedly change the risk profile of the plant from that which was licensed. We are not persuaded by the rationale provided by RES in favor of the first two-hour dose calculation. We agree with the position taken in the proposed final rule, and recommend-that the rule and the associated Regulatory Guides and SRP sections be issued. Sincerely,
- f. U T. S. Kress Chairman
REFERENCE:
Memorandum dated March 6,1996, from T. P. Speis, Office of Nuclear Regulatory Research, NRC, to J. T. Larkins, ACRS, transmitting Revisions to 10 CFR Part 100, Reactor Site Criteria, Revisions to 10 CFR Part 50, New Appendix S to Part 50 (Final Rules) and Associated Regulatory Guides and Standard Review Plan Sections
e 4 ATTACHMENT 4 DRAFT PUBLIC ANNOUNCEMENT l l
O O I l DRAFT PUBLIC ANN 0V_N. CEMENT The Nuclear Regulatory Commission (NRC) announced that it is issuing regulations to amend and to update the criteria used in decisions regarding l power reactor siting, including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. Existing reactor licensees would be unaffected by these changes. The revisions would allow the NRC to benefit from experience gained in the application of the procedures and methods used in the current regulation and to incorporate advancements in the earth sciences and earthquake engineering since the regulation was issued in j 1973. In addition, the regulations benefit from public comments received. i This rule primarily consists of two separate changes, namely, the source term , and dose considerations, and the seismic and earthquake engineering consideratiions of reactor siting. Basic reactor site criteria that have been j shown to be important to protecting public health and safety would be i incorporated inte the regulations, while source term and dose calculations ! i that apply primarily to plant design would be relocated. l ! In the seismic area, the rule would require thorough regional and site-specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would i be employed in plant design, whereas the Operating Basis Earthquake (OBE) would require a plant shutdown and inspection, were it to occur. l i l l l 1 i i I I
.iA A _,-_a 2-,.a ,_A m.'a.4.2J.. eel.-~%_a .A.--a.a2-L. ~Amam.,W...- m. Awa3 a.ha.3_.A.. 4m. . .--_e .
e 4 I l l l l i i ATTACHMENT 5 l l DRAFT CONGRESSIONAL LETTERS 1 i i r
g# G4 g 4 UNITED STATES 2 j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066tMo01 44 . . . . . ,o l l The Honorable Lauch Faircloth, Chairman l Subcomittee on Clean Air, Wetlands, Private l Property and Nuclear Safety Comittee on Environment and Public Works 1 United States Senate Washington, DC 20510
Dear Mr. Chairman:
Enclosed for the information of the Subcomittee are copies of a public announcement and a revision to Title 10 of the Code of Federal Regulations which is to be published in the Federal Reaister. The Nuclear Regulatory Commission 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. This rule would allow the NRC to benefit from experience gained in application of the procedures and methods contained in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In addition, this rule benefits from public coments received. This rule primarily consists of two separate changes, namely, the source term i and dose considerations, and the seismic and earthquake engineering i considerations of reactor siting. Basic reactor site criteria that have been shown to be important to protecting public health and safety would be incorporated into the regulations, while source term and dose calculations i that apply primarily to plant design would be relocated. In the seismic area, the rule would require thorough regional and site- , specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would j be employed in plant design, whereas the Operating Basis Earthquake (0BE) would require a plant shutdown and inspection, were it to occur. Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs Er. closures:
- 1. Public Announcement
- 2. Federal Register Notice cc: Senator Bob Graham l
. i p 44 g $ UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2066Mm01 o
% >.... /g The Honorable Dan Schaefer, Chairman Subcommittee on Energy and power Comittee on Comerce United States House of Representatives Washington, DC 20515
Dear Mr. Chairman:
1 Enclosed for the information of the Subcomittee are copies of a public l announcement and a revision to Title 10 of the Code of Federal Regulations j which is to be published in the Federal Reaister. ' The Nuclear Regulatory Comission 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. This rule would allow the NRC to benefit from experience gained in application of the procedures and methods contained in the current regulation I and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In addition, this rule benefits from public coments received. This rule primarily consists of two separate changes, namely, the source term l and dose considerations, and the seismic and earthquake engineering considerations of reactor siting. Basic reactor site criteria that have been shown to be important to protecting public health and safety would be incorporated into the regulations, while source term and dose calculations that apply primarily to plant design would be relocated. In the seismic area, the rule would require thorough regional and site-specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would be employed in plant design, whereas the Operating Basis Earthquake (0BE) would require a plant shutdown and inspection, were it to occur. Sincerely, Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosures:
- 1. Public Announcement .
- 2. Federal Register Notice cc: Representative Frank Pallone
1 i h 8
- 1 i
1 ATTACHMENT 6 DRAFT LETTERS TO THE SPE/.KER OF THE HOUSE OF REPRESENTATIVES, PRESIDENT OF THE SENATE, AND THE GENERAL ACCOUNTING OFFICE I i 1 1
-1
p ur
, og g i UNITED STATES -
;5 j
2 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4001 The Honorable Newt Gingrich Speaker of the United States House of Representatives l Washington, DC 20515 1
Dear Mr. Speaker:
Pursuant to Subtitle E of the Small Business Regulatory Enforcement Fairness Act of 1996, 5 U.S.C. 801, the Nuclear Regulatory Commission (NRC) is ! submitting a final rule that will update the criteria used in decisions ) regarding power reactor siting, including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. This rule would allow the NRC to benefit from experience gained in application of the
. ~ procedures and methods contained in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In addition, this rule benefits from public comments received.
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. Basic reactor site criteria that have been shown to be important to protacting public health and safety would be incorporated into the regulations, while source term and dose calculations that apply primarily to plant design would be relocated. In the seismic area, the rule would require thorough regional,and site-specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would be employed in plant design, whereas the Operating Basis Earthquake (OBE) would require a plant shutdown and inspection, were it to occur. l We have determined that this rule is not a " major rule" as defined in 5 U.S.C. l 804(2). We have confirmed this determination with the Office of Management and Budget. i Enclosed is a copy of the final rule, which is being transmitted to the Federal Register for publication. The Regulatory Flexibility Certification i.s l included in the final rule. Also enclosed is a copy of the Regulatory l Analysis for this final rule that contains the NRC's cost-benefit i determinations. This final rule is scheduled to become effective 30 days I after publication in the Federal Register. Sincerely,
- Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosures:
Final Rule l Regulatory Analysis i l
M fk & UNITED STATES p ! g j 2 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20656 4001 o 1
.... 9 :
The Honorable Al Gore : President of the United i States Senate ! Washington, DC 20510
Dear Mr. President:
Pursuant to Subtitle E of the Small Business Regulatory Enforcement Fairness Act of 1996, 5 U.S.C. 801, the Nuclear Regulatory Commission (NRC) is ! submitting a final rule that will update the criteria used in decisions l regarding power reactor siting, including geologic, seismic, and earthquake engineerin~g considerations for future nuclear power plants. This rule would allow the NRC to benefit from experience gained in application of the procedures and methods contained in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In addition, this rule benefits from public comments received. This rule primarily consists of two separate changes, nameiy, the source term and dose considerations, and the seismic and earthquake engineering considerations of reactor siting. Basic reactor site criteria that have been i shown to be important to protecting public health and safety would be ' incorporated into the regulations, while source term and dose calculations l that apply primarily to plant design would be relocated. In the seismic area, the rule would require thorough regional and site-specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would be employed in plant design, whereas the Operating Basis Earthquake (0BE) would require a plant shutdown and inspection, were it to occur.
^We have determined that this rule is not a " major rule" as defined in 5 U.S.C.
804(2). We have confirmed this determination with the Office of Management ' and Budget. Enclosed is a copy of the final rule, which is being transmitted to the Federal Register for publication. The Regulatory Flexibility Certification is l included in the final rule. Also enclosed is a copy of the Regulatory i Analysis for this final rule that contains the NRC's cost-benefit determinations. This final rule is scheduled to become effective 30 days after publication in the Federal Register. Sincerely, l l Dennis K. Rathbun, Director Office of Congressional Affairs
Enclosures:
Final Rule Regulatory Analysis l 1
~ . . - . . - - . - - - . -~ - _- - - . - - . - . - . - . - .--
muro y % UNITED STATES s* j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-0001 4 9 . . . . . ,o Mr. Robert P. Murphy General Counsel General Accounting Office Room 7175 441 G St., NW. Washington, DC 20548
Dear Mr. Murphy:
Pursuant to Subtitle E of the Small Business Regulatory Enforcement Fairness Act of 1996, 5 U.S.C. 801, the Nuclear Regulatory Commission (NRC) is l submitting a final rule that will update the criteria used in decisions regarding. power reactor siting, including geologic, seismic, and earthquake engineering considerations for future nuclear power plants. This rule would i allow the NRC to benefit from experience gained in application of the procedures and methods contained in the current regulation and to incorporate the rapid advancements in the earth sciences and earthquake engineering. In aNition, this rule benefits from public comments received. 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. Basic reactor site criteria that have been shown to be important to protecting public health and safety would be incorporated into the regulations, while source term and dose calculations i that apply primarily to plant design would be relocated. In the seismic area, the rule would require thorough regional and site-specific geoscience investigations. The Safe Shutdown Earthquake (SSE) would be employed in plant i design, whereas the Operating Basis Earthquake (0BE) would require a plant shutdown and inspection, were it to occur. We have determined that this rule is not a " major rule" as defined in 5 U.S.C. , 804(2). We have confirmed this determination with the Office of Management and Budget. Enclosed is a copy of the final rule, which is being transmitted to the Federal Register for publication. The Regulatory Flexibility Certification is l included in the final rule. Also enclosed is a copy of the Regulatory Analysis for this final rule that contains the NRC's cost-benefit determinations. This final rule is scheduled to become effective 30 days after publication in the Federal Register. l L Sincerely, Dennis K. Rathbun, Director , Office of Congressional Affairs
Enclosures:
Final Rule Regulatory Analysis
A +, ,a 6 --41.a.,---6d.- +Ln14 -,L,,a.--m.,-.ka: .4, m&+M. s aA 1 u a,a---a--4--es + 4.m_ aa ,, -- ---s.a- Asan- 4 a -AL4A<m .s.,a amA2, l l 4 t i .a i f ATTACHMENT 7 REGULATORY ANALYSIS i i l 1 l l 1
- e .
4 REGULATORY ANALYSIS REVISION OF 10 CFR PART 100 j AND 10 CFR PART 50 !- STATEMENT OF THE PROBLEM } This Regulatory Analysis covers two topics. First is the final rule revising 10
- CFR Part 100, " Reactor Site Criteria," for future plants. The second topic is
! a final rule codifying geologic and seismic siting factors for new plants. Both j- topics address the r.elocation to 10 CFR Part 50 plant design criteria from Part 100 and Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power i Plants," to 10 CFR Part 100. The first proposed revision to these regulations
- was published for public comment on October 20, 1992 (57 FR 47802). Due to the substantive nature of the changes, the Commission requested that all parts (10 2 CFR Parts 50 and 100), be reissued for public comment (Ref.1). The second
[ proposed revision to these regulations was published for public comment on 1 October 17, 1994 (59 FR 52555). ]
- - This regulatory analysis is presented in two parts, corresponding to the two i
s considerations stated above. ! Reactor Sitina Criteria (Nonseismic) The NRC's regulations in 10.CFR Part 100, " Reactor Site Criteria," present a , framework that guides the Commission in its evaluation of the suitability of 4 proposed sites for stationary power and testing reactors. The present criteria j regarding reactor siting were issued in April 1962. There were only a few small power reactors operating at that time. The present regulation requires that every reactor have an exclusion area that has no residents, although transient use is permitted. A low population zone immediately beyond the exclusion area is also required. The regulation recognizes the importance of accident considerations in reactor siting; hence, a key element in it is the determination of the size of the exclusion area via the postulation of' a large accidental fission product release within containment and the evaluation of the radiological consequences in terms of doses. Doses are calculated for two hypothetical-individuals, located at any point (generally, the closest point) on the exclusion area < boundary and at the outer radius of the low population zone, and are required to be within specified limits (25 rem to the whole body and 300 rem to the thyroid gland). In addition, the nearest population center, containing about 25,000 or more residents, must be no closer than one and one-third times the outer radius of the low population zone. The effect of these requirements is to set both individual and, to some extent, societal limits on dose (and implicitly on risk) without setting nur,erical criteria on the size of the exclusion area and low population zone. In practice the source term and dose calculations contained in 10 CFR 100 have influenced aspects of reactor design, such as containment leak rate and- performance of fission product cleanup systems such as sprays or filters, more than siting. Since the issuance of Part 100 in 1962, there have been significant changes and developments in power reactor technology. The nuclear power industry has developed and matured significantly. From the existence of a few small power plants, the industry has grown until there are presently about 110 power reactors in operation on 69 sites in the United States. Light-water commercial power reactors have accumulated about 2000 reactor-years of operating experience in the United States. Reactor power levels have also significantly increased. Early RA - 1
i plants typically had reactor power levels of about 150 megawatts thermal, whereas recently licensed plants have power. levels about 20 to 25 times greater. There has been increased development of and reliance upon fission product cleanup systems in modern plants to mitigate the consequences of postulated accidents.
< As a result, present nuclear power plants could be located at sites with a very small exclusion area and still meet the dose criteria of Part 100.
l There has also been an increased awareness and concern over potential nuclear ' accidents. In addition, there has been significant research on nuclear accidents including the factors leading to their initiation as well as accident ; phenomenology and progression. Although accident considerations have been of key importance in reactor siting from the very beginning, major developments in risk < assessment such as the issuance of the Reactor Safety Study (WASH-1400, Ref. 2), and NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants" (Ref. 3), as well as the occurrence of the Three Mile Island accident in 1979, and the accident at Unit 4 of the Chernobyl reactor in the Soviet Union in 1986, have' greatly increased awareness, knowledge, and concerns in this area. Since initial promulgation of Part 100 in 1962, the Commission has approved more than 90 sites for nuclear power plants and has had an opportunity to review a l number of others. As a result of these reviews, much experience has been gained regarding how siting factors influence and affect risk. l The substantial base of knowledge accumulated over the last 30 years on reactor siting, design, construction and operation reflect the fact that the major factors that determine public health and safety are the reactor design, construction and operation. Siting factors and criteria, however, are important in assuring that the radiological doses from normal operation and postulated accidents will be acceptably low, that natural phenomena and potential man-made hazards will be appropriately factored into the design of the plant, and that site characteristics are amenable to the development of adequate emergency plans to protect the public and adequate security measures to protect the plant. The Commission believes that the criteria for siting power reactors should provide basic site criteria that reflect the significant experience gained since the regulation was first issued in 1962. Seismic Sitina and Earthouake Enaineerina Criteria Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to 10 CFR Part 100, " Reactor Site Criteria," sets forth a framework that guides the staff in its evaluation of the adequacy of applicants' investigations of geologic and earthquake phenomena and proposed plant design para.neters. The issuance of Appendix A was an important step in establishing a definitive regulatory framework for dealing with earth science issues in the licensing of nuclear power plants. Appendix A contains the following statement:
"These criteria are based on the limited geophysical and geological information available to date concerning- faults and earthquake occurrence and effect. They will be revised as necessary when more complete information becomes available."
RA - 2
- l. .
The bases for Appendix A were established in the late 1960s and became effective December 13, 1973. Since then, with advances in the sciences of seismology and ; geology, along with the occurrence of some licensing issues not foreseen in the development of Appendix A, a number of significant difficulties have arisen in the application of this regulation. Specific problematic areas include the following:
- 1. In making geoscience assessments, there is a need for considerable latitude in judgment. This latitude in judgment is needed because -
l of limitations in data and geologic and seismic analyses, and l because of the rapid evolution taking place in the geosciences in
- terms of accumulating knowledge and in modifying concepts. This need was recognized when Appendix A was developed. However, having detailed geoscience assessments in Appendix A, a regulation, has created difficulty for applicants and the staff in terms of inhibiting the use of needed 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.
- 2. Various sections of Appendix A lack clarity and are subject to different interpretations and dispute. Also, some sections in the Appendix do not provide sufficient information for implementation.
As a result of being both overly detailed in some areas and not detailed enough in others, the Appendix has been the source of licensing delays and debate and has inhibited the use of some types i of analyses such as probabilistic seismic hazard analysis.
- 3. In other siting areas, such as hydrology, regulatory guidance has ,
been handled effectively through use of regulatory guides. Many problems encountered in implementing Appendix A could best be alleviated through the use of regulatory guides and a program for continuous updating. ;
- 4. The Operating Basis Earthquake (OBE) is associated with (i) the functionality of those features necessary for continued operation without undue risk to the health and safety of the public, (ii) an earthquake that could reasonably be expected to affect the plant site during the operating life of the plant, (iii) a minimum fraction of the Safe Shutdown Earthquake (SSE), and (iv) plant shutdown if vibratory ground motion is exceeded. These multi-aspects have resulted in seismic criteria that have led to overly stiff piping systems and excessive use of snubbers and supports which, in fact, could result in less reliable piping systems. Also, i regulatory guidance defining an exceedance of the OBE, and plant i shutdown or restart procedures have not been developed. Post '
earthquake evaluations are handled on an ad-hoc basis. ;
- 5. The stipulation in Appendix A that the SSE response spectra be defined at the foundation of the nuclear power plant structures has ,
l often led to confrontations with many in the engineering community i l who regard this stipulation as inconsistent with sound practice. l 1 RA - 3 l 1
.eo -m <
~ n OBJECTIVES Reactor Sitina Criteria (Nonseismic)
The objective of this regulatory action is to provide a stable regulatory basis for siting nuclear power plants by stating basic site criteria in Part 100 that reflects past experience, operational results, and research insights. This is accomplished by:
- a. providing basic site criteria reflecting past experience and )
importance to risk and
- b. relocating those requirements that apply to reactor design from Part 100 to Part 50. l The major, changes associated with the revision of the regulation are:
i
- 1. The regulatory action applies to applicants who apply for a construction or early site permit on or after the effective date of the final regulations. The current regulation will remain in place j
and be applicable to all licensees and applicants prior to the effective date of the final regulations. ) l
- 2. Part 100 states basic site criteria.
- 3. Source term and dose calculations are relocated to Part 50 )
consistent with the location of other design requirements in the l regulation. Since the revision to the regulation will not be a backfit, the licensing bases for existing nuclear power plants must remain in the regulation. Therefore, the revised regulation is designated as a new subpart to Part 100 for future plants while the current Part 100 is maintained for existing plants. Finally, in support of the above changes, Regulatory Guide 4.7 has been revised. Seismic Sitina and Earthauake Enaineerina Criteria The objectives of the regulatory action are to:
- 1. Provide a stable regulatory basis for seismic and geologic siting and applicable earthquake engineering design of future nuclear power plants that will avoid licensing delays due to unclear regulatory requirements;
- 2. Provide a flexible structure to permit consideration of new technical understandings; and
- 3. Have the revision to the regulation completed prior to the receipt of an early site application.
The major points associated with the revision of the regulation are: RA - 4
- 1. The regulatory action applies to applicants wh'o apply for an early site permit, design certification, or combined license (construction permit and operating license) pursuant to 10 CFR Part 52, or a construction permit or operating license pursuant to 10 CFR Part 50 on or after the effective date of the final regulation. However, for those operating license applicants and holders whose construction permit was issued prior to the effective date of the final regulation, the seismic and geologic siting and earthquake engineering criteria in Appendix A to 10 CFR Part 100 continues to apply. ;
1
- 2. Criteria not associated with the selection of the site or l establishment of the safe shutdown earthquake ground motion have been placed in Part 50. This action is consistent with the location of other design requirements in Part 50. ;
i Because the criteria presented in the final regulation does not apply to existing plants, th'e licensing bases for existing nuclear power plants must remain part of the regulations. Therefore, the revised criteria on seismic and geologic siting ~ is designated as a new s 100.23, " Geologic and Seismic Siting Factors," to 10 CFR Part 100 and is added to the existing body of regulations. Earthquake engineering criteria is located in 10 CFR Part 50 in a new Appendix
- S. Since Appendix S is not self executing, applicable sections of Part 50 (i.e.,
550.34,550.54) are revised to reference Appendix S. The rule makes conforming amendments to 10 CFR Parts 52 and 100. Finally, in support of the above changes, several regulatory guides and standard ' review plan sections are revised or developed as appropriate. ALTERNATIVES Reactor Sitina Criteria (Nonseismic) The alternatives considered included:
. No action (e.g., continue to use existing Part 100)
. Delete the existing Part 100 and replace it with an entirely new Part 100 that eliminates the dose calculation and specifies site criteria.
. Retain the existing Part 100 for current plants and add a new i
section to Part 100 for future plants that eliminates the dose calculation and specifies site criteria. The first alternative considered by the Commission was to continue using current : regulations for site suitability determinations. This is not considered an acceptable alternative. Accident source terms and dose calculations currently influence plant design requirements as well as siting. It is considered : desirable to state basic siting criteria which, through importance to risk, have i been shown to be key to assuring public health and safety. Further, significant ' advances in the earth sciences and in earthquake engineering, that deserve to be reflected in the regulations, have taken place since the promulgation of the , present regulation. RA - 5
1
- J Deletion of the existing regulation also is not considered an acceptable i
- alternative since it is the licensing bases for virtually all the operating
nuclear power plants and those in various stages of obtaining their operating l license. l l Therefore, the last option is the preferable course of action and is the option 1 4 evaluated further in this analyses. Seismic Sitina and Earthouake Enaineerina Criteria The first alternative considered by the Commission was to avoid initiating a i rulemaking proceeding. This is not an acceptable alternative. Althoi!gh the siting related issues associated with the current generation of nuclear power i plants are completed or nearing completion, there is a need to initiate the regulatory action in light of the current and future staff review of advanced reactor seismic design criteria. The current regulation has created difficulties i for applicants and the staff in terms of inhibiting flexibility in applying basic principles to new situations and using evolved methods of analysis in the licensing process. A second alternative considered was the deletion of the existing regulation (Appendix A to Part 100). This is not an acceptable alternative because these provisions form part of the licensing bases for many of the operating nuclear j power plants and others that are in various stages of obtaining their operating license. Also, geologic and seismic siting criteria are needed for future plants. Since there are problems with implementing the existing regulation (Appendix A to Part 100), the only satisfactory alternative is to revise the regulation. The 1 4 approach of establishing the revised requirements in a new Section 100.23 to Part 1 100 or Appendix S to Part 50 while retaining the existing regulation was chosen . as the best alternative. This approach is consistent with the current body of regulations; that is, requirements associated with seismology and geology, like , meteorology and hydrology, are contained within Part 100 not an appendix to Part 100. Similarly, detailed requirements associated with Part 50 are contained in appendices to Part 50 not within the sections of Part 50. 4 Finally, the following memoranda or reports provide further support for a revision to Appendix A to Part 100:
- 1. Staff Requirements Memorandum from Chilk to Taylor dated January 25, 1991,
Subject:
SECY-90-341 - Staff Study on Source Term Update and Decoupling Siting from Design (Ref. 4).
"The staff should further ensure that the revisions to Appendix A of Part 100 are available to support the time schedule shown in the paper [ Commission Briefing on Source Term Update and Decoupling Siting from Design (SECY-90-341), dated December
- 13,1990) for option 2, and are technically ;
supportable with the information that will 3 be available at the time the draft comes forward for Commission action." RA - 6
i I
- 2. Memorandum from Taylor to Beckjord dated September 6,1990,
Subject:
Revision of Appendix A,10 CFR Part 100, " Seismic and Geologic Siting Criteria for Nuclear Power Plants" (Ref. 5).
"I approve of your plan to begin work on the development of a revised regulation and this activity should be assigned a high i priority status." l
- 3. NUREG-0625, Siting Policy Task Force (Ref. 6).
" Revise Appendix A to 10 CFR Part 100 to better reflect the evolving technology in ,
assessing seismic hazards." :
- 4. NUREG-1061, " Report of the U.S. Nuclear Regulatory Commission Piping Review Committee," Vol 5, April 1985 (Ref. 7).
"The Committee recommends that o Rulemaking amending Appendix A to 10 CFR Part 100 be undertaken to permit decoupling of the OBE and SSE... ."
CONSEQUENCES
- a. Costs and Benefits _
e Benefits + Beactor Sitina Criteria (Nonseismic) The revision to Part 100 is beneficial to all. The industry and the public will benefit from a clearer, more uniform and consistent licensing process. Benefits to the industry, the public, and the NRC staff Will result from the following changes 1
- 1. Clear Statement of Basic Site Criteria. The revision to Part 100 provides basic site criteria with regard to acceptably low radiological consequences under normal operation and postulated accident conditions, assurance that natural phenomena as well as man-made hazards are factored into the plant design, and that the site is amenable to the development of adequate emergency plans and security measures. In addition, the criteria have been selected to be consistent with past experience and with the ,
quantitative health objectives in the NRC Safety Goal Policy. ;
- 2. Current Practices Will Be Reflected. The final regulations reflect industry design practices and the associated staff review procedures that have evolved since Part 100 was issued in 1962. An example of this is the requirement that man-made hazards from nearby industrial and ,
transportation facilities will be appropriately considered in the plant 1 design. Review of this area has been a part of the staff review for many years. Hence, the rule involves no substantive changes in this area. RA - 7
- - e- - r ==
- 3. Source Term and Dose Calculations. . The final rule relocates the use of a postulated source term and the calculation of radiological consequences to -
Part. 50 to reflect that these largely affect aspects of reactor design. ' The radiological- consequences are expressed in total effective dose equivalent (TEDE), which is consistent with usage in Part 20 and amenable with the use of a revised and updated source term consisting of nuclides in addition to the noble gases and iodine. :
- 4. Risk to the Public. The NRC Staff has generated a reduced set of source terms based on the NUREG-1150 (Ref. 3) analyses and the Independent Risk Assessment Plant. These source terms were used in the MELCOR Accident Consequences Code System (MACCS) for six reactor-containment designs. The results of these analyses indicate that the risk to the public is acceptably low and the quantitative health objectives (QHO) of the Commission's Safety Goal Policy are met for all plants up to 3800 MWt, the largest capacity plant considered in the analyses.
Seismic Sitina and Earthauake Enaineerina Criteria The revision of Appendix A to Part 100 is beneficial to all. The public will benefit from a clearer, more uniform and consistent licensing process subject to fewer interpretations. The NRC staff will benefit from improved regulatory I implementation (both technical and . legal), fewer interpretive debates, and l increased regulatory flexibility. Applicants will derive the same_ benefits in i addition to avoiding licensing delays because of unclear regulatory requirements. l l The regulatory action reflects changes intended to (1) benefit from the public comments associated with the first and second proposed revision of the current ] regulation, (2) benefit from the experience gained in applying the existing i regulation; (3) resolve interpretative questions; (4) provide needed regulatory I flexibility to incorporate state-of-the-art improvements in the geosciences and l earthquake engineering; (5) simplify the language to a more " plain English" text; and (6) acknowledge various internal staff and industry comments.
)
Benefits to applicants or NRC staff will result from the following changes:
- 1. Uncertainties and probabilistic methods. 'The new regulation (Section 100.23) 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 i 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. Therefore, methods other than the probabilistic methods, such as sensitivity analyses, may be adequate for some sites to account for uncertainties. 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, RA - 8
I l l Determine if information from geoscience investigations change probabilistic results, Conduct probabilistic seismic hazard analysis and determine ground motion level corresponding to the target exceedance probability Determine site-specific spectral shape and scale this shape to the ground motion level determined above, l NRC staff review of ground motion ' Update the data base and reassess probabilistic methods at ; least every ten years. l l Thus, the rule is anchored by the Comission Severe Accident Policy and requires thorough regional and site-specific geoscience investigations. In general, the approach reflects the comments of the U.S. utility industry. Results of the regional and site-specific investigations must be-considered in application 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) seismic hazard study, are essentially 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 to make the probabilistic analysis site-specific. 1 It is also necessary to incorporate local site geological factors such as stratigraphy and topography and to account for site-specific geotechnical properties in establishing the design basis ground motion. In order to incorporate local site factors and advances in ground motion attenuation models, ground motion estimates are determined using the procedures outlined in Standard Review Plan Section 2.5.2, Revision 3, " Vibratory Ground Motion." The NRC staff's review approach to evaluate an application is described in SRP Section 2.5.2. This review takes into account the information 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, the 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 assist in + assessing the fairly complex regional probabilistic modeling which incorporates multiple hypotheses and a multitude of parameters. ; Furthermore, this process should provide a clear basis for the staff's decisions and facilitate communication with nonexperts.
- 2. Reflect current design practices. The final regulations reflect industry design practices and the associated staff review procedures (for instance, the location of the control point for the seismic input) that have evolved since the initial regulation (Appendix A to Part 100) was issued in 1973. Many of these practices and RA - 9
procedures were incorporated into the revision of Standard Review Plan Sections 2.5.2, 3.7.1, 3.7.2, and 3.7.3 that are associated with the resolution of Unresolved Safety Issue (USI) A-40, " Seismic Design Criteria."
- 3. Clarify the multi-facets associated with the Operating Basis Earthquake (OBE). In the existing regulation, the OBE is associated with (1) the functionality of those features necessary for continued operation without undue risk to the health and safety of the public, (2) an earthquake that could reasonably be expected to affect the plant site during the operating life of the plant, (3) a minimum fraction of the Safe Shutdown Earthquake (SSE), and (4) plant shutdown if the vibratory ground motion is exceeded. In some cases, for instance, piping, the multi-facets of the OBE made it possible for the OBE to have more design significance than the SSE. The seismological basis, that is, the association of the OBE with a likelihood of occurrence has been removed from the regulation.
~ Other facets of the OBE, for instance, its value (percent of the SSE) and relationship with plant shutdown are discussed below. The functionality aspect of the OBE remains unchanged.
- 4. Value of the Operating Basis Earthquake Ground Motion (0BE) and required OBE analysis. The final regulation allows the value of the OBE to be set at (1) one-third or less of the SSE, where OBE requirements are satisfied without an explicit response or design analyses being performed, 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 5, Required Plant Shutdown), and second, the amount of analyses associated with the OBE. An applicant may determine that at one-third of the SSE level, the probability of exceeding. the OBE vibratory ground motion 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 performed 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. ) As stated above, it is determined that if an OBE of one-third of the SSE is used, the requirements of the OBE can be satisfied without the applicant performing 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 RA - 10 l
O,
- this position to seismic base isolation of buildings are discussed below. There is high confidence that, at this ground-motion 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, for future designs PRA insights will be used to support a margins-type assessment of seismic events. A PRA-based scismic margins analysis will consider sequence-level High Confidence, Low Probability of Failures (HCLPFs) and fragilities for all sequences leading to core damage or containment failures t.p 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 Analysis of External Events, memorandum from 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 OBE is associated with the design reqcirements, for example, the ultimate heat sink (see Regulatory Gutde 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 the final rule identifies existing guides that would .
- be revised technically to maintain the existing design philosophy. }
In SECY-93-087, " Policy, Technical, and Licensing Issues Pertaining . to Evolutionary and Advance Light-Water Reactor (ALWR) Designs," the l NRC. staff requested Commission approval on 42 technical and policy i issues pertaining to either evolutionary LWRs, passive LWRs, ' or both. The issue pertaining to the eliminati% of the OBE is designated I.M. .The NRC staff identified actions necessary for the design of structures, systems, and components when the OBE design requirement is eliminated. The staff clarified - that guidelines ; should be maintained to ensure the functionality of components, : equipment, and their supports. In addition, the staff clarified how !' certain design requirements are to be considered for buildings and l structures that are currently designed for the OBE, but not the SSE. , Also, the. NRC staff has evaluated the effect on safety of eliminating the OBE from the design load combinations for selected structures, systems, and componer.ts and has. developed proposed criteria for an analysis using only the SSE. Commission approval is i 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.
- 5. Guidance for required plant shutdown. The regulation treats plant ,
shutdown associated with vibratory ground motion exceeding the OBE or significant plant damage as a condition in every operating license. ' The shutdown requirement is a condition of the license (10 : CFR 50.54) .rather than a limiting. condition of operation -(10 CFR RA - 11
_ _ _ _ ~ __ _ _ . _ . . _ _ _ . _ _ _ _ ~ . _ _ _ . _ _ _ _ _ . _ . .. 50.36), because the necessary judgements associated with exceedance of the vibratory ground motion or significant plant damage can not be adequately characterized in a technical specification. A new paragraph, s50.54(ff) is added to the regulations to require plant shut down for licensees of nuclear power plants that comply 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. Regulatory Guide 1.166, " Pre-Earthquake Planning and Immediate Nuclear Power Plant Operator Post-Earthquake Actions,"
(Draft was DG-1034) has been developed to provide guidance acceptable to the NRC staff for determining whether or not vibratory ground motion exceeding the OBE or significant plant damage had occurred and nuclear power plant shut down is required. The guidance is based on criteria developed by the Electric Power Research Institute (EPRI). Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by a Seismic Event," (Draft was DG-
. 1035) has been developed to provide guidelines that are acceptable to the NRC staff for performing inspections and tests of a nuclear power plant equipment and structures prior to plant restart. This
. guidance is also based on EPRI reports.
- 6. Reduced level of detail. The level of detail presented in the final regulations has been limited to general guidance. The final regulations identify and establish basic requirements. Detailed guidance, that is, the procedures acceptable to the NRC for meeting the requirements, has been removed and placed in Regulatory Guide, 1.165, " Identification and Characterization of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motions," (Draft was DG-1032).
- 7. Provide greater flexibility. The regulations provide a flexible structure that will permit the consideration of new technical understandings and state-of-the-art advancements since the detailed guidance has been removed from the regulation and placed into regulatory guides.
- 8. Clarify interpretations. Changes have been made to the seismic and geologic siting criteria to resolve past questions of interpreta-tion. As an example, the-definitions and required investigations sections of the final regulation has been significantly changed to eliminate or modify phrases that were more applicable to only the western United States. .
- 9. Clarify text. The regulations use more explicit terminology. For j instance, the Safe Shutdown Earthquake (SSE) and Operating Basis -
Earthquake are now referenced as the Safe Shutdown Earthquake Ground , Motion (SSE) and the Operating Basis Earthquake Ground Motion (OBE). j In' addition, appropriate changes within the text-highlight that the i SSE used as the design basis is not associated with a single > earthquake but characterized by free-field ground motion response spectra. l RA - 12
o . Costs Reactor Sitina Criteria (Nonseismic) , The costs associated with the revised regulations are subdivided into two categories; the first is associated with siting criteria modifications (Part 100), the second is associated with (Part 50) modifications. Part 100 -, The overall cost impact associated with revising the s.iting criteria aspects of ' the regulation are neutral. Important factors in this regard are:
- 1. Nearby Industrial and Transportation Facilities. This area of review is incorporated into the regulations as one of the basic site criteria. It has been a part of the staff review for many years. l
~
The rule involves no substantive changes in this area and would merely codify what has been staff practice for a number of years. 2.- Feasibility of Carrying out Protective Actions. The rule requires . that the site characteristics be amenable to the development of ; adequate emergency plans. Emergency plans are currently required in 10 CFR 50.47. Hence, this rule imposes no new requirements but requires early assurance of emergency planning feasibility as part of the site review process, possibly reducing time and costs at the OL or COL stage by avoiding licensing delays. The cost impact associated with this revision is neutral. The revision is expected to increase time and costs for site approval - but should significantly reduce time and costs at the OL or COL ' stage by avoiding licensing delays.
- 3. Feasibility of Developing Adequate Security Measures. The rule requires that the site characteristics be such that adequate security measures to protect the plant can be developed. Security measures are c9rrently required in 10 CFR Part 73. Hence, this rule imposes no new requirements but requires early assurance of the :
feasibility of developing security measures as part of the site , review process, possibly reducing time and costs at the OL or COL -
- stage by avoiding licensing delays. ,
The cost impact associated with this revision is neutral. The revision is expected to increase time and costs for site approval but should significantly reduce time and costs at the OL or COL t stage by avoiding licensing delays. Part 50 The overall cost impact associated with revising the reactor licensing aspects of the regulation are neutral because the source term and dose calculations have always been required under Part 100 for site suitability but are now required under Part 50 and used in evaluating plant features. RA - 13
Seismic Sitina and Earthauake Enaineerina Criteria -The costs associated with the regulations are subdivided into two categories; the first is associated with the geosciences and site investigations (Section 100.23), the second is associated with earthquake engineering (Appendi.t S to Part 50). 10 CFR 100.23 4 The overall cost impact associated with the geosciences and site investigation aspects of the regulation as compared to Appendix A of Part 100 are slightly increased in some areas but reduced overall because of anticipated improvement in the licensing process. Specific examples include: l
)
- 1. Reduced Licensing Delays. The licensing process is enhanced because information needed for the staff review can be incorporated in the
~ safety analysis reports at the time of docketing instead of later through staff questions and applicant responses. I
- 2. Probabilistic Evaluations. Probabilistic evaluations to determine vibratory ground motion, surface tectonic deformation, and seismically induced floods and water waves reflect to some extent what is already current staff practice. In particular, probabi-listic hazard analyses have been used to determine the probability of exceeding the Safe Shutdown Earthquake Ground Motion at the plant site. However, the overall use of probabilistic evaluations as suggested in Regulatory Guide 1.165, " Identification and Characterization of Seismic Son ces and Determination of the Safe i Shutdown Earthquake Ground Mot',ons," is new but should not have a i significant cost impact. Computer codes to . perform the probabilistic analyses are available. An applicant would input the site coordinates - and local site effects (current requirement) to obtain the probabilistic hazard data. It is estimated that these analyses can be performed within a few days.
Appendix S to Part 50 The overall cost impact associated with the earthquake engineering aspects of the regulation 'are neutral or reduced. Specific examples include:
- 1. Reduced OBE Analysis. The response analyses associated with the Operating Basis Earthquake Ground Motion (0BE) is eliminated if the applicant sets the OBE at one-third of the Safe Shutdown Earthquake Ground Motion (SSE). Selecting an OBE value greater than one-third of the SSE does not increase the analytical effort above current requirements.
- 2. Control Point Location. Changing the location of the control point (the point at which the vibratory ground motion is applied) from the foundation level to the free-field does not affect costs. The following discussion from Section 2.1.1.4 of NUREG-1233 (pages 13 and 14) is applicable:
"A number of recent plants were designed to the 1975 Standard Review Plan requirements RA - 14 '
i, . a i . which specified the free-field motion at soil-structure the free-surface for i i interaction analysis. During the operating license (OL) review, the implementation of 3 i i the current position of input motion at the- l foundation level in the free field resulted t in a modification of some structural floor l- beams of seismic Category I structures at i one plant. No hardware changes resulted at .
- other . plants. (Note that the staff's
4 investigation was limited. to the Safe ; shutdown systems and structures that housed ; them, and allowance was made for tested ( strength values in some cases.)" ! ! 3. Seismic Instrumentation. Although the seismic instrumentation ! ~ requirements are different (only time-history accelerographs instead i l of time-history accelerographs, response spectrum recorders and peak
- accelerographs), the cost is essentially the same as that associated j with operating plants; there are fewer instruments required. The
- maintenance and calibration costs with the new solid-state seismic ,
, instrumentation are less than that associated with the current ! ) instrumentation. The processing of instrumentation data will be : done at the site, thereby reducing the potential for prolonged plant i
; shutdown while data are being evaluated. In general, the ability to i i expeditiously assess the effects of the earthquake on the plant will '
l- save both staff and licensee resources.
- 4. Post-Earthquake Activities. In preparation of postaarthquake l l activities, it is recommended that the licensee inspect and
- base-line. certain structures, equipment and piping. Base line '
inspections would aid in differentiating between pre *xisting )
- conditions at the nuclear power plant and earthquake related damage. !
i The structures, equipment and piping selected for these inspections r
- are comprised of those routinely examined by plant operators during
- normal plant walkdowns and inspections. After an earthquake, plant 1 operators familiar with the plant would walkdown and visually j inspect accessible areas of the plant. Unnecessary plant shutdowns l would be avoided since the pre-earthquake condition of equipment and '
structures (for example, physical appearance, leak rates, vibration levels) would be known. This approach has been submitted to the NRC staff for approval by the Nuclear Management and Resources Council ' (NUMARC) (now the Nuclear Energy Institute (NEI)) and is documented
- in an Electric Power Research Report, EPRI NP-6695, " Guidelines for
! Nuclear Power Plant Response to an Earthquake." The associated cost
- impact is minimal and recommended by industry.
. IMPACTS i ! a. Other NRC Proarams i { None for the Nonseismic siting criteria. i i Although Appendix A to 10 CFR Part 100 is titled " Seismic and Geologic l Siting Criteria for Nuclear Power Plants," it is also referenced in two
- M - 15 i
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 Dispos" ion of Tailings or Waste Produced by the Extraction or Concentration .f Source Material from Ores Processed Primarily for Their Source Many tal Content," Section I, Criterion 4(e), and (2) Part 72, " Licensing Ramirements for the Independent Storage of Spent Nuclear Fuel and High-Leve? Radioactive Waste," Paragraphs (a)(2)(b) and (a)(2)(f)(1) of 572.102. In conjunction with the second proposed revision to the regulations the Department of Energy (Office of Civilian Radioactive Waste Management), requested that an explicit statement be added to the Statement of Consideration as to whether or not s 100.23 applies to the Mined Geologic Disposal System (MGDS) and a Monitored Retrievable Storage (MRS) facility. DOE provided the following documentation: (1) NRC has noted in NUREG-1451,
" Staff Technical Position on Investigations to Identify Fault Displacement Hazards and Seismic Hazards at a Geologic Repository," that Appendix A to 10 CFR Part 100 does not apply to a geologic repository; (2) NUREG-1451 also notes that the contemplated revisions to Part 100 would also not be applicable to a geologic repository; and (3) 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.
In response, the staff stated that 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 5 100.23 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.
- b. Other Government Aaencies Since the siting and licensing of nuclear power plants is carried out solely by NRC staff, no impact is projected for other government agencies.
- c. Constraints None.
DECISION RATIONALE Reactor Sitina Criteria (Nonseismic) The major considerations that have guided the Commission in this revisien to the reactor site' criteria are as follows:
- 1. The criteria will assure a low risk for individuals as well as for society in general, even in the event of severe but unlikely reactor accidents. The criteria are consistent with the Commission Safety Goal Policy with respect to the risk of both prompt and latent cancer fatalities. In addition, the Commission has examined severe accident risks associated with possible land contamination or property damage in the event of significaet releases of long-lived RA - 16
radioactive species, such as cesium. Siting away from densely ; populated centers is expected to result in a low likelihood of significant offsite contamination of densely populated areas.
- 2. The criteria will anure that man-made activities as well as natural events associated witi; the site location are identified and used in matching a design with the site. ;
- 3. The criteria will assure that site characteristics are such that ;
~
adequate emergency plans can be developed to protect the public.
- 4. The criteria will assure that site characteristics are such that adequate security measures to protect the plant can be developed. ;
- 5. The criteria will explicitly state the Commission's policy that 1 reactors should be sited away from densely populated centers. ;
The~ revisions reflect current staff practice. The revised regulations will s not reduce risk, but would improve the description in the regulations of current staff practice in licensing. , Seismic Sitina and Earthauake Enaineerina Criteria i The recommendations to revise the existing regulation (Appendix A to 10 CFR Pa*t l 100) by adding sections for future applications pertaining to the geosciences aid t site investigations (s 100.23) and earthquake engineering (Appendix S to Part 5)) are baseo primarily on qualitative rather than quantitative or probabilistic ! (i.e., core. damage frequency reduction) arguments. The staff's evaluation augments the regulatory analysis associated with the implementation of Unresolved 1 Safety Issue (USI) A-40, " Seismic Design Criteria" (NUREG-1233, Ref. 8). USI A-40 was implemented in August 1989 through the revision of Standard Review Plan Sections 3.7.1, " Seismic Design Parameters," 3.7.2, " Seismic System Analysis," 3.7.3, " Seismic Subsystem Analysis," and 2.5.2, " Vibratory Ground Motion." The staff's conclusion is that for operating reactor and operating license i applicants, the final regulations have little effect on risk. Operating plants j generally have been, and will be, seismically upgraded by plant-specific actions ' such as implementation of the Systematic Evaluation Program (SEP), the implementation of Generic Letter 88-20, Supplement 4, " Individual Plant Examinations of External Events (IPEEE) for Severe Accident Vulnerabilities," the implementation of USI A-46, " Verification of Seismic Adequacy of Equipment in Operating Plants," and NRC Bulletin programs. Therefore, this regulatory action is applicable only to applicants who apply for an early site permit, design certification, combined license, construction permit or operating license on or after the effective date of the final regulations. No overall increases in costs are expected in implementing the regulations for applicants for early site permits, design certifications, combined licenses, construction permits or operating license. In addition, the regulations will reduce delays in the licensing process because-information needed for the staff i review can be incorporated in the safety analysis reports at the time of l docketing instead of later through staff questions and applicant responses. Therefore, the staff proposes that all new applicants be required to comply with the revised regulations. RA - 17
I Current Reaulatory Action The current regulatory action consists of the following:
- 1. Revisions to $50.2, 550.8, $50.34, 550.54, and $52.17.
- 2. Revisions to $100.1, s100.2, 5100.3, and $100.8.
- 3. Add Subpart B, s100.20, 5100.21, and $100.23.
- 4. Add a new Appendix S to Part 50 Earthquake Engineering Criteria for Nuclear Power Plants
- 5. Issue new Regulatory Guides:
- a. Regulatory Guide 1.165, " Identification and Characterization-
~ of Seismic Sources and Determination of Safe Shutdown Earthquake Ground Motions,"-(Draft was DG-1032)
- b. Regulatory Guide 1.166, " Pre.-Earthquake Planning and Immediate Nuclear Power Plant Operator Post 4arthquake Actions," (Draft was DG-1034)
- c. Regulatory Guide 1.167, " Restart of a Nuclear Power Plant Shut Down by-a Seismic Event," (Draft was DG-1035)
- 6. Issue Revised Regulatory Guides:
- a. Regulatory Guide 4.7, Revision 2, " General Site Suitability Criteria for Nuclear Power Stations," (Draft was DG-4003)
- b. Regulatory Guide 1.12, Revision 2, " Nuclear Power Plant Instrumentation for Earthquakes," (Draft was DG-1033)
- 7. Issue Revised Standard Review Plan Sections:
2.5.1, Basic Geologic and Seismic Information. 2.5.2, Vibratory Ground Motion. 2.5.3, Surface Faulting. Future Reaulatory Action Several existing regulatory guides will be revised to incorporate editorial changes or. maintain the existing design or analysis philosophy. These guides will be issued subsequent to the publication of the final regulations that would implement this action. The following regulatory guides will be revised to incorporate editorial changes The type of changes contemplated would be to reference new paragraphs in Appendix B to Part 100 or Appendix S to Part 50: 1, 1.57, " Design Limits and Loading Combinations for Metal Primary Reactor Containment System Components" RA - 18
- 2. 1.59, " Design 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, " Bases for Plugging Degraded PWR Steam Generator Tubes"
- 8. 1.122, " Development of Floor Design Response Spectra for Seismic Design of Floor-Supported Equipment or Components" The following regulatory guides will be revised to maintain existing design or analysis philosophy. For example, the types of changes contemplated would be to change OBE to a fraction of the SSE.
- 1. 1.27, " Ultimate Heat Sink for Nuclear Power Plants"
- 2. 1.100, " Seismic Qualification of Electric and Mechanical Equipment for Nuclear Power Plants"
- 3. 1.124, " Service Limits and Loading Combinations for Class 1 Linear-Type Component Supports"
- 4. 1.130, " Service Limits and Loading Combinations for Class 1 Plate-and-Shell-Type Component Supports"
- 5. 1.132, " Site Investigations for Foundations of Nuclear Power Plants"
- 6. 1.138, " Laboratory Investigations of Soils for Engineering Analysis and Design of Nuclear Power Plants"
- 7. 1.142, " Safety-Related Concrete Structures for Nuclear Power Plants (Other than Reactor Vessels and Containments)"
- 8. 1.143, " Design Guidance for Radioactive Waste Management Systems, .
Structures, and Components Installed in Light-Water-Cooled Nuclear ' Power Plants" i If substantive changes are made during the revisions, the applicable guides will be issued for public comment as draft guides. , l i IMPLEMENTATION l This regulatory action is applicable only to applicants that apply for an early ) RA - 19
7 site permit, design certification, combined license, con'struction permit, or operating license on or after the effective date of the final regulations. For those operating license applicants and holders whose construction permit was ! issued prior to the effective date of the final regulation, the seismic and I geologic siting and earthquake engineering criteria in Appendix A to Part 100 continues to apply. 6 1 l I i RA - 20 l
l I REFERENCES
- 1. Memorandum from Samuel J. Chilk to James M. Taylor,
Subject:
SECY-94-017 - Options with Regard to Revising 10 CFR Part 100, Reactor Site Criteria, March 28, l 1994. l
- 2. U.S. Nuclear Regulatory Commission, " Reactor Safety Study-An Assessment of Risks in U.S. Commercial Nuclear Power plants," NUREG-75/014 (WASH-1400),
December 1975. ! 3. U.S. Nuclear Regulatory Commission, " Severe Accident Risks: An Assessment for l Five U.S. Nuclear Power Plants," NUREG-1150, December 1990.
- 4. Staff Requirements Memorandum from S.J. Chilk to J.M. Taylor, Subject SECY l 341, January 25, 1991.
- 5. Memorandum from J.M. Taylor to E.S. Beckjord, Subject Revision of Appendix A, 10 CFR Pait 100, September 6,1990. !
l
- 6. U.S. Nuclear Regulatory Comission, " Report of the Siting Policy Task Force," l NUREG-0625, August 1979. !
- 7. U.S. Nuclear Regulatory Comission, " Report of the U.S. Nuclear Regulatory Commission Piping Review Comittee," NUREG-1061, Volume 5, April 1985.
- 8. S.K. Shaukat and N.C. Chokshi, " Regulatory Analysis for USI A-40, ' Seismic Design Criteria,'" NUREG-1233, U.S. Nuclear Regulatory Comission, September 1989. 1
- 9. Electric Power Research Institute, " Guidelines for Nuclear Plant Response to l an Earthquake," NP-6695, December 1989.
1 I I l
- i RA - 21
p., ..,a a. a - a= - .- - - - - ~ -----s - a.- -- s a ~~. 1-- -- - -Lk 2 L ss--n1 ---""2<- r A~ ~ - 0 i ATTACHMENT 8 ENVIRONMENTAL ASSESSMENT l l l 1
ENVIRONMENTAL ASSESSMENT AND FINDING OF NO SIGNIFICANT IMPACT l REVISION OF 10 CFR PART 100. AND 10 CFR PART 50 The Nuclear Regulatory Commission is amending its regulations to update the reactor siting criteria, seismic and geologic siting criteria, and earthquake ' engineering criteria for nuclear- power plants. The first proposed revision to I these regulations was published for public comment on October 20, 1992 (57 FR l 47802). Due to the substantive nature of the changes, the Commission requested that all parts (10-CFR Parts 50 and 100, and Appendix A to 10 CFR Part 100) be reissued for public comment. The second proposed revision to these regulations was published for public comment on October 17, 1994 (59 FR 52255). The nonseismic and seismic areas are discussed separately. Identification of Action Reactor Sitino Criteria (Nonseismic) 10 CFR Part 100, " Reactor-Site Criteria," originally issued in April 1962, is revised. The revision will apply to applicants who-apply for site approval on er after the effective date of the final regulation. Since the revision to the regulation will not be a backfit, the bases for existing nuclear power plants must remain in the same regulation. Therefore, the revised regulation on siting is designated Subpart B of 10 CFR Part 100; the existing regulation is designated Subpart A of 10 CFR Part 100. Criteria not associated with site selection are relocated into Part 50 consistent with the location of other design requirements in the regulation. Hence, source term and dose calculations are relocated to Part 50.
.The rule ~ states basic site criteria including the need for the site characteristics to be such that radiological doses from both normal operation as well as postulated accidents are acceptably low, that natural phenomena and man-made' hazards must be appropriately factored into the design of the plant, that the site characteristics must be amenable to the development of emergency plans to protect the public and security measures to protect the plant. Reactor sites should also to be located away from very densely populated centers, and that i
areas of low population density are, generally, preferred. I Seismic Sitino and Earthauake Encineerino Criteria Appendix A, " Seismic and Geologic Siting Criteria for Nuclear Power Plants," to l 10 CFR Part 100, " Reactor Site Criteria," was originally issued as a proposed rule on November 25,1971 (36 FR 22601); published as a final rule on November 13, 1973 (38 FR 31279); and became effective on December 13, 1973. There have been two amendments to Appendix A to 10 CFR Part 100. The first amendment, ;
-issued November 27,1973 (38 FR 32575), corrected the final rule by adding the ,
legend under the diagram. The second amendment resulted from a petition for ! rulemaking (PRM 100-1) requesting that an opinion interpreting and clarifying Appendix A with respect to the determination of the Safe Shutdown Earthquake be issued. A notice of filing of the petition was published on May 14, 1975 (40 FR 20983). The substance of the petitioner's proposal was accepted and published as an immediately effective final rule on January 10, 1977 (42 FR 2052). EA - 1
! . c l
The amendment applies to applicants who apply for an early site permit, design certification, combined license, construction permit, or operating license on or . after the effective date of the final regulation. However, for those operating license applicants and holders whose construction permit was issued prior to the effective date of the regulation, the seismic and geologic siting and earthquake engineering criteria in Appendix A to 10 CFR Part 100 continues to apply. Because the revised criteria presented in the regulation will not'be applied to
- existing plants,- the licensing bases for existing nuclear power plants must remain part of the regulations. Therefore, the revised criteria on seismic and geologic siting is designated as a new Section 100.23, " Geologic and seismic siting factors," to 10 CFR Part 100, " Reactor Site Criteria," and has been added ,
- to the existing _ body of-regulations.
Criteria not associated with site selection or establishment of the Safe Shutdown i Earthquake Ground Motion (SSE) are placed in 10 CFR.Part 50. This action is consistent with the location of other design requirements in Part 50. Hence, earthquake engineer _ing criteria are located in Appendix S to 10 CFR Part 50, "Earthquak'e Engineering Criteria for Nuclear Power Plants." The regulatory action incorporates changes that are intended to (1) benefit from the experience gained in applying the existing regulation, (2) resolve ' interpretative questions, (3) provide needed regulatory flexibility to incorporate improvements in the geosciences and earthquake engineering, and (4) ; simplify the language to a more " plain English" text. 3 Need for the Action Reactor Sitina Criteria (Nonseismic) Since its initial promulgation in 1962, the Commission has approved more than 90 sites for nuclear power plants and has had an opportunity to review a number of others. As a result of these reviews, much experience has been gained regarding. the' site factors that influence risk and their range of acceptability. ; Additionally, there has also been increased awareness, concern and significant research on potential nuclear accidents. Although accident considerations have been of key importance in reactor siting from the very beginning, major developments in risk assessment such as the issuance of the Reactor Safety Study (WAS4-1400) in 1975, and the issuance of NUREG-1150, " Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants," in December 1990, as well as the occurrence of the Three Mile Island accident in 1979, and the Chernobyl accident in-the Soviet Union in 1986, have greatly increased awareness,- knowledge, and concerns in this area. The substantial base of knowledge accumulated over the last 30 years on reactor design, construction and operation reflect the fact that the major factors that determine public health and safety are the reactor design, construction and operation. Siting factors and criteria, however, are important in assuring that the radiological doses from normal operation and postulated accidents will be acceptably low, that natura1' phenomena and potential man-made hazards will be - appropriately factored into the design of the pl ant, and that site characteristics are amenable to the development of adequate emergency plans to EA - 2
o . protect the public and adequate security measves to protect the plant. The Commission believes that the criteria for siting power reactors should provide basic site criteria that reflect the significant experience learned since the regulation was first issued in 1962. Seismic Sitina and Earthauake Enaineerina Criteria The experience gained in the application of the procedures and methods set forth in the current regulation and the rapid advancement in the earth sciences and earthquake engineering have made it necessary to update the 1973 criteria. Environmental Impacts of the ""..QB Reactor Sitina Criteria (Nonseismic) Subpart B 'to Part 100 contains the considerations that will guide the Commission in its evaluation of the suitability of a proposed site for nuclear power plants after the effective date of the final regulation. The revision to Part 50 contains the engineering considerations for evaluation of the suitability of the i plant design. The amendment to 10 CFR Part 100 reflects current licensing l practice and does not change the radiological environmental impact. Stated differently, the regulatory actions for future siting applications (10 CFR Part 100 Subpart B) are based on maintaining about the level of risk of radiological releases as in the regulation (10 CFR Part 100, Subpart A) they replace. Seismic Sitina and Earthauake Enaineerina Criteria Section 100.23 to 10 CFR Par'. 100 contains the seismic and geologic considerations that guides the Comission in its' evaluation of the suitability of sites proposed for nuclear power plants and the suitability of the nuclear power plant design bases established in consideration of the seismic and geologic characteristics of the proposed sites. Appendix S to 10 CFR Part 50 contains the earthquake engineering considerations that guides the Commission in its evaluation of the suitability of the piant design bases. The revision of Appendix A to 10 CFR Part 100 as stated an Section 100.23 to 10 CFR Part 100 and Appendix S to 10 CFR Part 50 reflect current licensing practice in earthquake engineering and enhanced current staff practice in seismic and geologic siting through the use of probabilistic evaluations or other methods, such as sensitivity analyses, where applicable. The target exceedance probability is set by examining the design bases of more recently licensed nuclear power plants. Therefore, the radiological environmental impact offsite will not change. Stated differently, the regulatory actions (Section 100.23 to Part 100 and Appendix S l to Part 50) are specifically based on maintaining the present level of risk of l radiological releases, thus having zero effect compared to the regulation l (Appendix A to Part 100) they replace. 1 Onsite occupational radiation exposure associated with inspection and maintenance I will not change. These activities are principally associated with baseline l inspections of structures, equipment, and piping and maintenance of seismic instrumentation. 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 EA - 3
o inspections. Routine maintenance of seismic instrumentation ensures its operability during earthquakes. The location of the seismic instrumentation is similar to that in the existing nuclear power plants. In addition, the regulatory guide pertaining to seismic instrumentation (Regulatory Guide 1.12, Revision 2, " Nuclear Power Plant Instrumentation for Earthquakes") specifically cites occupational radiation exposure as a consideration in selecting the location of the instruments. > The amendments do not affect non-radiological plant effluents and have no other environmental impact. Therefore, the Commission concludes that there are no significant non-radiological environmental impacts associated with the amendments to the regulations.- Alternatives to the Action As required by Section 102(2)(E) of NEPA (42 U.S.C.A. 4332(2)(E)), the staff has considered' possible alternatives to the proposed action. The first alternative considered by the Commission was to avoid initiating a rulemaking proceeding. This is not an acceptable alternative. Present accident source terms and dose calculations presently influence plant design requirements rather than siting. It is considered desirable to be able to state basic site criteria which, through importance to risk, have been shown to be key to assuring public health and safety. Further, significant advances in the earth sciences and in earthquake engineering, that deserve to be reflected in the regulations, have taken place since the promulgation of the present regulation. A second ali.ernative considered was deletion of the existing regulation. This is not ~ an acceptable alternative because these provisions form the licensing bases for almost all operating nuclear power plants. For the seismic siting and ecrthquake engineering areas, another alternative considered was replacement of the entire regulation with a regulatory guide. , This is not acceptable because a regulatory guide is non-mandatory. The staff believes that there could be an increase in the risk of radiation exposure to the public if the siting and earthqucke engineering criteria were nonmandatory. The approach of- establishing new sections of the regulations for revised requirements while retaining the existing regulations was chosen as the best alternative. The public will benefit from a clearer, more uniform and consistent licensing process subject to fewer interpretations. The NRC staff will benefit from improved implementation (both technical and legal) of the regulations, fewer interpretive debates, and increased regulatory flexibility. Applicants will derive the same benefits in addition to avoiding licensing delays caused by unclear regulatory requirements. Adopting revised siting and engineering criteria would increase the efficiency of regulatory actions. Alternative Use of Resources No alternative use of resources was considered. 1 EA - 4 I
Aaencies and Persons Consulted Reactor Sitina Criteria (Nonseismic) The NRC staff developed the enclosed rulemaking recommendations. No outside agencies or consultants were used .in developing this rulemaking package. However, the rulemaking reflects the extensive public comments received during the proposed revisions. In addition, several public meetings were held to inform industry of the staff's efforts in revising the siting criteria. The NRC staff also obtained advice from the NRC Advisory Committee on Reactor Safeguards. Seismic Sitina and Earthauake Enaineerina Criteria During the development of the proposed regulations and supporting regulatory guides, the NRC staff had several public meetings with interested industry groups, principally, the Nuclear Energy Institute (NEI) (previously the Nuclear Management and Resources Council (NUMARC)) and the Electric Power Research Institute "(EPRI) . The NRC staff also obtained advice from the NRC Advisory Committee on Reactor Safeguards and comments from the U.S. Geological Survey 't (USGS) staff. As a proposed rule, the regulations were released for public comment to encourage participation from the public and various organizations in the development of the regulations. For example, comments received from the public on the first and second proposed revision of the regulations were considered in the development of the final regulations. Findina of No Sionificant Imoact The Commission has determined under the National Environmental Policy Act of 1969, as amended, that the amendments to 10 CFR Parts 50 and 100 that relocate dose calculation requirements, specify siting criteria (population, seismic, and geologic), and specify earthquake engineering criteria for nuclear power plants do not have a significant effect on the quality of the human environment and that an environmental impact statement is not required. This determination is based on the following:
- 1. The amendments to the regulations -largely reflect current practice, consistent with the staff's evaluation of applicant's safety analysis reports at the time of docketing, applicant's responses to staff initiated questions, and the results of research in the earth sciences and seismic engineering.
- 2. The foregoing environmental assessment.
- 3. The q Jiitative, deterministic, and probabilistic assessments pertaining to seismic events in NUREG-1070, NUREG-1233, and NUREG-1407 (References 1 through 3, respectively).
- 4. The Policy Statement on Severe Reactor Accidents Regarding Future Designs
- and Existing Plants, published August 8,1985 (50 FR 32138), affirming the Commission's belief that a new design for a nuclear power plant can be shown to be acceptable for severe accident concerns if the criteria and
- procedural requirements cited in 50 FR 32138 are met.
EA - 5 t
- 5. Commission approval, with modification, of the staff recommendation pertaining to site-specific Probabilistic Risk Assessments and analyses of ,
external events. As stated in Reference 4: "PRA insights will be used to support a margins-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 containx.snt failures up to approximately one and two-thirds the ground motion acceleration of the Design Basis SSE." References
- 1. "NRC Policy on Future Reactor Designs, Decisions on Severe Accident Issues in Nuclear Power Plant Regulation," NUREG-1070, July 1985.
- 2. "Regula, tory Analysis for USI A-40, " Seismic Design Criteria" Final Report," .
NVREG-1233, September 1989. l
- 3. " Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities, Final Report," I NUREG-1407, June 1991.
- 4. 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. 1 EA - 6
4 li t I 1 i ATTACHMENT 9
~
DRAFT REGULATORY GUIDE DG-4004 ;
- (GENERAL SITE SUITABILITY CRITERIA) i l
/
[ U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REGULATORY RESEARCH February 1995 Division 4 i l Task DG-4004 t i
/ DRAFT REGULATORY GUIDE
Contact:
L. Soffer (301)415-6574 4 DRAFT REGULATORY GUIDE DG-4004 i (The Second Proposed Revision 2 to Regulatory GE j (Previously issued as DG-4003) GENERAL SITE SUITAB l
^
CRITERIA FOR NUCLEAR ER IONS k c o
- t k.
r, : N l
<4 -
This regulatory guide is being issued in draft form to involve the p@lic in the early stages of the development of a regulatory position in this area. It hee not received complete staff review and does not represent en offidel NRC staff position. Public commente are being solicited on the draft guide (includn0 eny implementatHm schedde) and its sesoaisted re0detary onelysis or valuerempact statement. Conenents should be accompermed by appropriate supportm0 date. Written commente may be esmrtted to the Rules Reinew end Cirectives Branch OFIPS, Office of Admanistration, U.S. Nucieer Ragdetary Commmeion, Weehmeton, DC 20555. Copies of commente received may be exemmed et the NRC PubEc Document Room,2120 L Street NW., Weehmeton, DC. Comments will km most helpful y if received by May 12, 1995. Requeste for single copies of draft guides (wisch may be reproduced) or for piecement on en automatic detnbution list for single copies of futwo guides in specific dvisione should be made in writing to the U.S. Nuclear Regdetary Comnwesion, Weehmston, DC 20555, Attention: Office of Administration. Distribution and Mei! Services Section.
. . l I
l l i 1 TABLE OF CONTENTS l 2 A. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . I ; 5 ) 3 B. DISCUSSION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 4 GEOLOGY / SEISMOLOGY ........................ 5 l 5 ATMOSPHERIC EXTREMES AND DISPERSION . . . . . . . . . . . . . . . .
.............. 8 I 6 EXCLUSION AREA AND LOW POPULATION ZONE 9
7 POPULATION CONSIDERATIONS . . . . . . . . . . . . . . . . . . . . . i EMERGENCY PLANNING ........................ 9 8 SECURITY PLANS .......................... 10 l 9 10 HYDROLOGY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 11 i 11 Flooding .......................... 12 Water Availability ..................... 11 13 . Wat er Qu al i ty . . . . . . . . . . . . . . . . . . . . . . . . 12 14 INDUSTRIAL, MILITARY, AND TRANSPORTATION FACILITIES . . . . . . . . 13 ECOLOGICAL SYSTEMS AND BIOTA ................... 16 15 , 16 LAND USE AND AESTHETICS . . . . . . . . . . . . . . . . . . . . . . 19 17 SOCI0 ECONOMICS. . . . . . . . . . . . . . . . . . . . . . . . . . . 21 18 NOISE . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 , 19 C. REGULATORY POSITION . . . . . . . . . . . . . . . . . . . . . . . . . 22
- 1. GEOLOGY / SEISMOLOGY . . . . . . . . . . . . . . . . . . . . . . . 22
~ 20
.............. 23 21 2. ATMOSPHERIC EXTREMES AND DISPERSION 22 3. EXCLUSION AREA AND LOW POPULATION ZONE . . . . . . . . . . . . . 23
................... 24 g 23 4. POPULATION CONSIDERATIONS 24 5. EMERGENCY PLANNING . . . . . . . . . . . . . . . . . . . . . . . 25
'- 25 6. SECURITY PLANS . . . . . . . . . . . . . . . . . . . . . . . . . 26 27 26 7. HYDROLOGY ...........................
7.1 Flooding ........................ 27 27 28 7.2 Water Availability .............~...... 27 29 7.3 Water Quality . . . . . . . . . . . . . . . . . . . . . . 27 30 7.4 Fission Product Retention and Transport . . . . . . . . . 28
- 8. INDUSTRIAL, MILITARY, AND TRANSPORTATION FACILITIES ...... 28 31 32 9. ECOLOGICAL SYSTEMS AND BIOTA . . . . . . . . . . . . . . . . . . 30 33 10. LAND USE AND AESTHETICS . . . . . . . . . . . . . . . . . . . . 32 34 11. SOCIOECONOMICS ........................ 32 33 35 12. NOISE . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
IMPLEMENTATION ........................... 33 36 D. 37 APPENDIX A - SAFETY-RELATED SITE CONSIDERATIONS FOR ASSESSING SITE
.............. A-1 38 SUITABILITY FOR NUCLEAR POWER STATIONS 39 APPENDIX B -ENVIRONMENTAL CONSIDERATIONS FOR ASSESSING B-1 40 SITES FOR NUCLEAR POWER STATIONS .................
41 DRAFT REGULATORY ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . RA-1 G i
j
'1 A. INTRODUCTION 2 The Atomic Energy Act of 1954 (42 U.S.C. 2011 et seq.), as amended, and i 3 the Energy Reorganization Act of 1974 places on the Neclear Regulatory
- 4 Commission (NRC) the responsibility for the licensing aid regulation of pri- i vate nuclear facilities from the standpoint of public health and safety. Part 5
6 100, " Reactor Site Criteria," of Title 10 of the Code of Federal Regulations 7 requires that the population density; use of the site environs, including 8 proximity to man-made hazards; and the physical characteristics of the site, 9 including seismology, meteorology, geology, and hydrology, be taken into l 10 account in determining the acceptability of a site for a nuclear power 71 reactor. Seismic and geologic site criteria for nuclear power plants are 12 provided in Appendix A and in a proposed Section 100.23 of 10 CFR Part 100 (59 13 FR 52255). Appendix A to 10 CFR Part 50 establishes minimum requirements for 14 the principal design criteria for water-cooled nuclear power plants; a number 15 of these criteria are directly related to site characteristics as well as to 16 events and conditions outside the nuclear power unit. O 17 18 The National Environmental Policy Act of 1969 (NEPA) (42 U.S.C. 4321 et seq.) as amended, implemented by Executive Order 11514 and the Council on 19 Environmental Quality's Guidelines of November 28, 1978 (43 FR 55990), 20 requires that all agencies of the Federal Government prepare detailed 21 environmental statements on proposed major Federal actions that can siptifi-22 cantly affect the quality of the human environment. A principal objective of 23 NEPA is to require the Federal agency to consider, in its decision-making 24 process, the environmental impacts of each proposed major action and the 25 available alternative actions, including alternative sites. 26 Part 51, " Environmental Protection Regulations for Domestic Licensing
, 27 and Related Regulatory Functions," of Title 10 of the Code of Federal 28 Regulations sets forth the NRC's policy and procedures for the preparation and 29 processing of environmental impact statements and related documents pursuant 30 to Section 102(2)(C) of the NEPA.
31 The limitations on the Commission's authority and responsibility 32 pursuant to the NEPA imposed by the Clean Water Act [ Federal Water Pollution 33 Control Act (FWPCA)] (33 U.S.C.1251 et seq.), as amended, are addressed in 34 the Policy Statement Regarding Implementation of Certain NRC and EPA Responsi-35 bilities published in the Federal Reaister on December 31, 1975 (40 FR 60115). 1 l
.l l
1 This guide discusses the major site characteristics related to public 2 health and safety and environmental issues that the NRC staff considers in 3 determining the suitability of sites for light-water-cooled (LWR) nuclear l 4 power stations.' The guidelines may be used by applicants in identifying 5 suitable candidate sites for nuclear power stations. The decision that a O 6 station may be built on a specific candidate site is based on a detailed 7 evaluation of the proposed site-plant combination and a cost-benefit analysis 8 comparing it with alternative site-plant combinations as discussed in 9 Regulatory Guide 4.2, " Preparation of Environmental Reports for Nuclear Power 10 Stations,."' 11 Chapter 9 of Regulatory Guide 4.2 discusses the selection of a site from 12 among alternative sites; the applicant should present its site-plant selection 13 process as the consequence of an analysis of alternatives whose environmental 14 costs and benefits were evaluated and compared and then weighed against those 15 of the proposed facility. 16 This guide is intended to assist applicants in the initial stage of 17 selecting potential sites for a nuclear power station. Each site that appears 18 to be compatible with the general criteria discussed in this guide will have 19 to be examined in greater detail before it can be considered to be a "candi-20 date" site, i.e., one of the group of sites that are to be considered in 21 selecting a " proposed" or " preferred" site.' 22 *For the purpose of this guide, nuclear power station refers to the nuclear 23 reactor unit or units, nuclear steam supply, electric generating units, 24 auxiliary systems including the cooling system and structures such as docks 25 that are located on a given site, and any new electrical transmission towers 26 and lines erected in connection with the facilities. 27
- Copies are available for inspection or copying for a fee from the NRC Public 28 Document Room at 2120 L Street NW., Washington, DC; the PDR's mailing address 29 is Mail Stop LL-6, Washington, DC 20555; telephone (202)634-3273; fax 30 (202)634-3343. Copies may be purchased at current rates from the U.S.
31 Government Printing Office, P.O. Box 37082, Washington, DC 20402-9328 32 (telephone (202)512-2249); or from the National Technical Information Service 33 by writing NTIS at 5285 Port Royal Road, Springfield, VA 22161. 34 *See Chapter 9 of Regulatory Guide 4.2 for a discussion of site selection 35 procedures. The " proposed" site submitted by an applicant for a construction 36 permit is that site chosen from a number of " candidate" sites the applicant 37 prefers and on which the applicant proposes to construct a nuclear power 38 Jtation. i
4 1 This guide should be used only in the initial stage of site selection 2 because it does not provide detailed guidance on the various relevant factors : 3 and format for ranking the relative suitability or desirability of possible 4 sites. This guide provides a general set of safety and environmental criteria 5 that the NRC staff has found to be valuable in assessing candidate site 6 identification in specific licensing cases. l 7 The information needed to evaluate potential sites at this initial stage ; 8 of site selection is assumed to be limited to information that is obtainable
~
9 from published reports, public records, public and private agencies, and 10 individuals knowledgeable about the locality of a potential site. Although in 11 somi cases the applicants may have conducted on-the-spot investigations, it is 12 assumed here that these investigations would be limited to reconnaissance-type 13 surveys at this stage in the site selection process. 14 The safety issues discussed include geologic / seismic, hydrologic, and , II meteorological characteristics of proposed sites; exclusion area and low l 16 population zone; population considerations as they relate to protecting the 17 general public from the potential hazards of serious accidents; potential 18 effects on a station from accidents associated with nearby industrial, 19 transportation, and military facilities; emergency planning; and security 20 plans. The environmental issues discussed concern potential impacts from the 21 construction and operation of nuclear power stations on ecological systems, 22 water use, land use, the atmosphere, aesthetics, and socioeconomics. l 23 This guide does not discuss details of the engineering designs required 24 to ensure the compatibility of the nuclear station and the site or the 25 detailed information required for the preparation of the safety analysis and i 26 environmental reports. In addition, nuclear power reactor site suitability as 27 it may be affected by the Commission's materials safeguards for nuclear power 28 plants is not addressed in this guide. 29
~
A significant commitment of time and resources may be required to select 30 a suitable site for a nuclear power station, including safety and environ-31 mental considerations. Site selection involves consideration of public health 32 and safety, engineering and design, economics, institutional requirements,. 33 environmental impacts, and other factors. The potential impacts of the 34 construction and operation of nuclear power stations on the physical and i 3
. .l 1
1 biological environment and on social, cultural, and economic features
- 2 (including environmental justice) are usually similar to the potential impacts 3 of any major industrial facility, but nuclear power stations are unique in the 1 4 degree to which potential impacts of the environment on their safety must be 5 considered. The safety requirements are primary determinants of the 6 suitability of a site for nuclear power stations, but considerations of 7 environmental impacts are also important and need to be evaluated.
8 In the site selection process, coordination between applicants for 9 nuclear power stations and various Federal, State, and local agencies will be 10 useful in identifying potential problem areas. 11 Appendices A and B of this guide sunenarize the important safety-related 12 and environmental considerations for assessing the site suitability of nuclear 13 power stations. 14 Regulatory guides are issued to describe and make avsilable to the 15 public such information as methods acceptable to the NRC staff for 16 implementing specific parts of the Commission's regulations, techniques used 17 by the staff in evaluating specific problem.s or postulated accidents, and i 18 guidance to applicants. Regulatory guides are not substitutes for regulations, 19 and compliance with regulatory guides is not required. Regulatory guides are 20 issued in draft form for public comment to involve the public in the early 21 stages of developing regulatory positions. Draft regulatory guides have not 22 received complete staff review and do not represent official NRC staff 23 positions. 24 Any information collection activities mentioned in this draft regulatory 25 guide are contained as requirements in the proposed amendments to 10 CFR Part 26 50 that would provide the regulatory basis for this guide. The proposed 27 amendments have been submitted to the Office of Management and Budget for 28 clearance that may be appropriate under the Paperwork Reduction Act. Such 29 clearance, if obtained, would also apply to any information collection 30 activities mentioned in this guide. 31 ' Biological and physical environment includes geology, geomorphology, , 32 surface and groundwater hydrology, climatology, air quality, limnology, water 33 quality, fisheries, wildlife, and vegetation. Social and cultural features 34 include scenic resources, recreation resources, archeological and historical 35 resources, and community resources, including land use patterns. 4 1 1
B. DISCUSSION Q 1 V 2 GE0 LOGY / SEISM 0 LOGY 3 Nuclear power stations must be designed to prevent the loss of safety-4 related functions. Generally, the most restrictive safety-related site char-5 acteristics considered in determining the suitability of a site are surface 6 faulting, potential ground motion and foundation conditions' (including 7 liquefaction, subsidence, and landslide potential), and seismically induced 8 floods. Criteria that describe the nature of the investigations required to 9 obta1n the geologic and seismic data necessary to determine site suitability 10 have been set forth in a proposed amendment to 10 CFR Part 100, " Reactor Site 11 Criteria," in Section 100.23, " Geologic and Seismic Siting Factors" (59 FR 12 52255). Safety-related site characteristics are identified in Section 2.5 of 13 Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports 14 for Nuclear Power Plants,"* and Regulatory Guide 1.59, " Design Basis Floods 15 for Nuclear Power Plants."' In addition to geologic and seismic evaluation for assessing seismically induced flooding potential, Section 2.4 of Q 16 V 17 Regulatory Guide 1.70 and Regulatory Guide 1.59 describe hydrologic criteria, 18 including coincident flood events that should be considered. 1p ATMOSPHERIC EXTREMES AND DISPERSION l 20 The potential effect of natural atmospheric extremes (e.g., tornadoes' 21 and exceptional icing conditions') on the safety-related structures of a f nuclear station must be considered. However, the atmospheric extremes that 22 23 may occur at a site are not normally critical in determining the suitability i 1 24 'W.J. Hall, N.M. Newmark, and A.J. Hendron, Jr., " Classification, Engineering 25 Properties and Field Exploration of Soils, Intact Rock and In Situ Rock 26 Masses," WASH-1301, May 1974, outlines some of the procedures used to evaluate 27 site foundation properties. Copies are available for inspection or copying 28 for a fee from the NRC Public Document Room at 2120 L Street NW., Washington, 29 DC; the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555; 30 telephone (202)634-3273; fax (202)634-3343. 31 ' Refer to Regulatory Guide 1.76, " Design Basis Tornado for Nuclear Power Plants." 32 ' Refer to Section 2.4.7 of Regulatory Guide 1.70. 5
1 of a site because safety-related structures, systems, and components can be l 2 designed to withstand cost atmospheric extremes. 3 The atmospheric characteristics at a site are an important consideration 4 in eval'uating the dispersion of radioactive effluents from both postulated 5 accidents and routine releases in gaseous effluents.' In addition to 6 meeting the NRC requirements for the dispersion of airborne radioactive 7 material, the station must meet State and Federal requirements of the Clean !
- 8. Air Act (42 U.S.C. 7401 et seq.) as amended. This is unlikely to be an 9 important consideration for nuclear power station siting unless (1) a site is 10 in an area where exieting air quality is near or exceeds standards, (2) there 11 .is a potential for interaction of the cooling system plume with a plume 12 containing noxious or toxic substances from a nearby facility, or (3) the
~13 auxiliary generators are expected to operate routinely.
14 The atmospheric data necessary for assessment of the potential 15 ' dispersion of radioactive material are described in Regulatory Guide 1.23, 16 "Onsite Meteorological Programs."' 17 In_ the evaluation of potential sites, onsite meteorological monitoring l 18 can determine if the atmospheric conditions at a site are adequately repre- , 19 sented by the available atmospheric data for the area. Canyons or deep 20 valleys frequently have atmospheric variables that are substantially different 21 from those variables measured for the general region. Other topographical 22- features such as hills, mountain ranges, and lake or ocean shorelines can 23 affect the local atmospheric conditions at a site and may cause the dispersion 24 characteristics at the site to be less favorable than those in the general 25 area or_ region. More stringent design or effluent objectives may be required 26 in such cases. 27 ' Radiation doses associated with routine releases of airborne radioactive 28 material must be kept "as low as is reasonably achievable" (ALARA) [see 10 29 CFR 20.1101(b)]. 30 The requirements for design objectives for equipment to control releases 31 of radioactive material in effluents from nuclear power reactors are set forth 32' in the proposed 50.34(a). 33 Further, 10 CFR 50.36a provides that, in order to keep power reactor 34 effluent releases ALARA, each license authorizing operation of such a facility 35 will include technical specifications regarding the establishment of effluent 36 control equipment and reporting of actual releases. 37 Appendix I to 10 CFR Part 50 provides numerical guidance for design I 38 objectives and technical specification requirements for limiting conditions of 39 operation for light-water-cooled nuclear power plants. 6
1 1 While it is the concentration of radioactive materials in the atmosphere i 01 2 3 at' any distance from the point of release, x(Ci/m'), that must be controlled, the ratio y/0, where Q(Ci/sec) is the rate of release of radioactive materials 4 from the source, has become a comonly evaluated term because it depends only 5 on atmospheric variables and distance from the source. 6 If the dispersion of radioactive material released following a design 7 basis accident is insufficient at the boundary of the exclusion area (see the 8 following section, " Exclusion Area and Low Population Zone") and the outer 9 boundary of the low population zone, the plant design would not satisfy the requirements proposed in Paragraph 50.34(a)(1). In this case, the design of 10 11 the st'ation would be required to include appropriate and adequate compensating 12 engineered safety features. In addition, meteorological conditions are to be 13 determined for use (1) in the environmental report required in 10 CFR Part 51, 14 (2) for comparison to the assumptions in the Probabilistic Risk Assessment 15 (PRA) for a certified plant design (if such a design is to located at the 16 site) or to the assumptions in the site-specific PRA for a custom plant at the 17 site, and (3) for verification of the criteria specified in the Design Control Document for a certified plant design. N 18 (V 19 Local fogging and icing can result from water vapor discharged into the 20 atmosphere from cooling towers, lakes, canals, or spray ponds, but can gen-21 erally be acceptably mitigated by station design and operational practices. 22 However, some sites have the potential for severe fogging or icing because of 23 local atmospheric conditions. For example, areas of unusually high moisture 24 content that are protected from large-scale airflow patterns are most likely 25 to experience these conditions. The impacts are generally of greatest poten-26 tial importance relative to transportation or electrical transmission systems 27 in the vicinity of a site. 28 A cooling system designed with special consideration for reducing drift 29 may be required because of the sensitivity of the natural vegetation or the 30 crops in the vicinity of the site to damage from airborne salt particles. The 31 vulnerability of existing industries or other facilities in the vicinity of 32 the site to corrosion by drift from cooling tower or spray system drift should 33 be considered. Not only are the amount, direction, and distance of the drift 34 from the cooling system important, but the salt concentration above the natural background salt deposition at the site is also important in assessing
/N 35
\ 36 drift effects. None of these considerations are critical in evaluating the 7
\
I suitability of a site, but they could result in special cooling system design 2 requirements or in the need for a larger site to confine the effects of drift 3 within the site boundary. The environmental effects of salt drift are most 4 severe where saline water or water with high mineral content is used for 5 condenser cooling. 6 Cooling towers produce cloudlike plumes that vary in size and altitude 7 depending on the atmospheric conditions. The plumes are often a few miles in 8 length before becoming dissipated, but the plumes themselves or their shadows 9 could have aesthetic impacts. Visible plumes emitted from cooling towers in 10 the vicinity of airports could cause a hazard to aviation. 11 EXCLUSION AREA AND LOW POPULATION ZONE 12 A reactor licensee is required by 10 CFR Part 100 to designate an j 13 exclusion area and to have authority to determine all activities within that 14 area, including removal of personnel and property. In selecting a site for a l 15 nuclear power station, it is necessary to provide for an exclusion area in 16 which the applicant has such authority. Transportation corridors such as ! 17 highways, railroads, and waterways are permitted to traverse the exclusion 18 area provided (1) these are not so close to the facility as to interfere with 19 normal operation of the facility and (2) appropriate and effective arrange-20 ments are made to control traffic on the highway, railroad, or waterway in 21 case of emergency to protect the public health and safety. 22 The prcposed Section 50.34(ii)(D)(1) would require that the exclusion 23 area must be of such a size that an individual assumed to be located at any 24 point on its boundary would not receive a radiation dose in excess of 25 rem 25 total effective dose equivalent (TEDE) over any 2-hour period following a 26 postulated accidental fission product release into the containment. The 27 required exclusion area size involves consideration of the atmospheric 28 characteristics of the site as well as plant design. 29 A reactor licensee is also required by 10 CFR Part 100 to designate an 30 area immediately beyond the exclusion area as a low population zone (LPZ). The 31 size of the LPZ must be such that the distance to the boundary of the nearest 32 densely populated center containing more than about 25,000 residents must be 33 at least one and one-third times the distance from the reactor to the outer 34 boundary of the LPZ. The boundary of the population center should be 8
determined upon consideration of population distribution, not political 02 1 3 boundaries. The proposed Section 50.34 would require that the LPZ be of such a size 4 that an individual located on its outer radius for the course of the 5 postulated accident (assumed to be 30 days) would not receive a radiation dose 6 in excess of 25 rem TEDE. The size of the LPZ depends upon atmospheric 7 dispersion characteristics and population characteristics of the site as well 8 as aspects of plant design. 9 POPULATION CONSIDERATIONS 10 As stated in 10 CFR Part 100, reactors should be located away from very 11 densely populated centers and areas of low population density are, generally, 12 preferred. Part 100 also states that, in determining the acceptability of a 13 particular site located away from a very densely populated center but not in 14 an area of low density, consideration will be given to safety, environmental, 15 economic, or other factors that may result in the site being found acceptable. Locating reactors away from densely populated centers is part of the 9 16 17 NRC's defense-in-depth philosophy and facilitates emergency planning and 18 preparedness as well as reducing potential doses and property damage in the 19 event of a severe accident. The numerical values given in this guide (see 20 Regulatory Position 4, " Population Considerations") are generally consistent 21 with past NRC practice and reflect consideration of severe accidents as well 22 as the demographic and geographic conditions of the United States.
. '23 EMERGENCY PLANNING 24 The proposed Section 100.21(g) would require that " site characteristics 25 must be such that adequate plans to take protective actions for members of the 26 public in the event of emergency can be developed."
I 27 Additionally,10 CFR 50.47(a)(1) requires reasonable assurance that 28 adequate protection can and will be taken in the event of a radiological 29 emergency before an operating license for a nuclear powe' plant can be issued. 30 Adequate plans must be developed for two areas or Emergency Planning Zones (EPIs). As stated in 10 CFR 50.47, the plume exposure path'.a.y EPZ for nuclear 4 31 32 power plants generally consists of an area about 16 km (10 mi) in radius, and 9
l I the ingestion pathway EPZ generally consists of an area about 80 km (50 mi) in 2 radi u's. 3 The exact size and configuration of the EPZs should be determined in 4 relation to local emergency response needs and capabilities as they are 5 affected by such conditions as demography, topography, land characteristics,
! 6 access routes, and jurisdictional boundaries.
7 SECURITY PLANS 8 The proposed Section 100.21 would require that " site characteristics 9 must be such that adequate security plans and measures can be developed." 10 Physical protection requirements for nuclear power plants as well as special 11 nuclear materials are described in 10 CFR Part 73. Security plans and 12 measures are important to prevent plant damage and possible radiological 13 consequences to members of the public as a result of acts of sabotage. 34 Based on experience and analysis, the NRC staff has found that a l 15 distance of about 110 meters (360 feet) to any vital structure or vital i 16 equipment generally would provide sufficient space to satisfy security 17 measures specified in 10 CFR 73.55 (e.g., protected area barriers, detection 18 equipment, isolation zones, vehicle barriers). Since the distance to the 19 nearest exclusion area boundary is considerably greater than 110 meters (360 20 feet), the site characteristics are not nonna11y limiting with regard to the 21 ability to develop adequate security plans. 22 A possible exception occurs if the exclusion area is traversed by a 23 highway, railroad, or waterway. Traversal of such routes through the 24 exclusion area is permitted, provided they are not so close that they I 25 interfere with normal operations of the facility, and provided appropriate and ) 26 effective arrangements have been made to control traffic on such routes in I 27 case of emergency. If a transportation route passes closer than about 110 28 meters (360 feet) to a vital structure or vital equipment, special measures or 29 analyses may be needed to show that adequate security plans can be developed. 4 4 10
1 HYDROLOGY 2 Floodine 3 Criteria for evaluation of seismically induced floods are provided in 4 Section 100.23 to 10 CFR Part 100. Regulatory Guide 1.59, " Design Basis 5 Floods for Nuclear Power Plants,"* describes an acceptable method of 6 determining the design basis floods for sites along streams or rivers and 7, discusses the phenomena producing comparable design basis floods for coasta!, 8 estuary, and Great Lakes sites. The effects of a probable maximum flood (as 9 de'fi,ned in Regulatory Guide 1.59), seiche, surge, or seismically induced flood 10 such as might be caused by dam failures or tsunami on station safsty functions 11 can generally be controlled by engineering design or protection of the safety-12 related structures, systems, and components identified in Regulatory Guide 13 1.29, " Seismic Design Classification."' For some river valleys, flood plains, ' 14 or areas along coastlines, there may not be sufficient information to make the 15 evaluations needed to satisfy the criteria for seismically induced flooding. In such cases, study of the potential for dam failure, river blockage, or O'16 17 diversion in the river system or distantly and locally generated sea waves may be needed to determine the suitability of a site. In lieu of detailed 18 19 investigations, Regulatory Guide 1.59 and Section 2.4 of Regulatory Guide 1.70 20 present acceptable analytical techniques for evaluating seismically induced 21 flooding, j 22 Water Availability l 23 Nuclear power stations require reliable sources of water for steam l 24 condensation, service water, emergency core cooling system, and other func-t 25 tions. Where water is in short supply, the recirculation of the hot cooling I 26 water through cooling towers, artificial ponds, or impoundments has been 27 p'racticed. 28 Water requirements for nuclear power plants are that sufficient water be 29 available for cooling during plant operation and normal shutdown, for the 30 ultimate heat sink, and for fire protection. The limitations imposed by existing laws or allocation policies govern the use and consumption of cooling 9 31 32 water at potential sites for normal operation. Regulatory Guide 1.27, 11 1 I
t i. i 1 " Ultimate Heat Sink for Nuclear Power Plants,"' provides guidance on water f 2 supply for the ultimate heat sink and discusses the safety requirements. l 3 Consumption of water may necessitate an evaluat' ion of existing and future ;
- 4 water uses in the area to ensure adequate water supply during droughts for l 5 both station operation and other water users (i.e., nuclear power station
- 6 requirements versus public water supply). Regulatory agencies should be
} 7 consulted to avoid potential conflicts. ; j 8 Where required by law, demonstration of a request for certification of l 9 the rights to withdraw or consume water and an indication that the request is l ) 10 consistent with appropriate State and regional programs and policies is to be l )~ 11 provided as part of the application for a construction permit or operating l
'12 license. l l'
13 The availability of essential water during periods of low flow or low 14 water level is an important initial consideration for identifying potential i { 15 sites on rivers, small shallow lakes, or along coastlines. Both the frequency
; 16 and duration of low flow or low-level periods should be determined from the ! 17 historical record and, if the cooling water is to be drawn from impoundments, 18 from projected operating practices.
I ! 19 Water Quality , 20 Thermal and chemical effluents discharged to navigable streams are 21 governed by the Federal Water Pollution Control. Act (FWPCA) (33 U.S.C.1251 et
- 22 seq.) as amended, 40 CFR Part 122, 40 CFR Part 423, and State water quality j 23 standards. The applicant should also determine whether there are other i 24 regulations that are current at the time sites are under considerat' ion.
! 25 Section 401(a)(1) of the FWPCA requires, in part, that any applicant for an ! 26 NRC construction permit or combined license (combined construction permit and i 27 operating license) for a nuclear power station provide to the NRC i 28- certification from the State that any discharge will comply with applicable , { 29 effluent limitations and other water pollution control requirements. In the I i 30 absence of such certification, no construction permit or combined license can
! 31 be issued by NRC unless the requirement is waived by the State or the State l ) 32 fails to act within a reasonable period of time. A National Pollution i 33 Discharge Elimination System (NPDES) permit to discharge effluents to navi- g l 34 gable streams pursuant to Section 402 of the FWPCA may be required for a 1 i
12
1 i nuclear power station to operate in compliance with the Act, but it is not a 1 rx 1 s 2 'rerequisite to an NRC construction permit or operating license. p 3 Evaluations of the dispersion and dilution capabilities and potential 4 contamination pathways of the ground-water environment under operatin'g and 5 accident conditions with respect to present and future users are required. l 6 Potential radiological and nonradiological contaminants of ground water should l 7 be evaluated. The suitability of sites for a specific plant design in areas ! 8 with a complex ground-water hydrology or of sites located over aquifers that 9 are or may be used by large populations for domestic or industrial water 10 supplies or for irrigation water can only be determined after reliable 11 assessments have been made of the potential impacts of the reactor on the 12 ground water. Accordingly,10 CFR Part 100 requires that site environmental 13 parameters, which include hydrological and meteorological characteristics, be 14 characterized and used in or compared to those used in the plant PRA and 15 environmental analysis. 16 Although management of the quality of surface waters is important, water I l 17 quality is not generally a determining factor in assessing the suitability of 18 a site since adequate design alternatives can be developed to meet FWPCA 9 19 20 requirements and the Commission's regulations implementing NEPA. The following are examples of potential environmental effects of station 21 construction and operation that must be assessed: physical and chemical 22 environmental alterations in habitats of important species, including plant-23 induced rapid changes in environmental conditions; changes in normal current 24 direction or velocity of the cooling water source and receiving water; 25 scouring and siltation resulting from construction and cooling water intake 26 and discharge; alterations resulting froni dredging and spoil disposal; and 27 interference with shoreline processes.
. 28 INDUSTRIAL. MILITARY. AND TRANSPORTATION FACILITIES I
29 Accidents at present or projected nearby industrial, military, and 30 transportation facilities may affect the safety of a nuclear power station l 31 (see Section 2.2 of Regulatory Guide 1.70). The proposed Section 100.21 would 32 require that, " Potential hazards associated with nearby transportation routes, 33 industrial and military facilities must be evaluated and site parameters
, 6. 34 established such that potential hazards from such routes and facilities will 13
e I pose no undue risk to the type of facility proposed to be located at the 2 site." 3 Accidents at nearby industrial facilities such as chemical plants, 4 refineries, mining and quarrying operations, oil or gas wells, or gas and 5 petroleum product storage installations may produce missiles, shock waves, 6 flammable vapor clouds, toxic chemicals, or incendiary fragments. These may 7 affect the station itself or the station operators in a way that jeopardizes 8 the safety of the station. 9 Accidents at nearby military facilities, such as munitions storage areas 10 and ordnance test ranges, may threaten station safety. An otherwise 11 unacceptable site may be shown to be acceptable if the cognizant military 12 organization agrees to change the installation or mode of operation to reduce 13 the likelihood or severity of potential accidents involving the nuclear
- 4 station to an acceptable level.
15 An accident during the transport of hazardous materials (e.g., by air, 16 waterway, railroad, highway, or pipeline) near a nuclear power plant may gen-17 erate shock waves, missiles, and toxic or corrosive gases that can affect the 18 safe operation of the station. The consequences of the accident will depend 19 on the proximity of the transportation facility to the site, the nature and , 20 maximum quantity of the hazardous material per shipment, and the layout of the 21 nuclear station. 22 Airports are transportation facilities that pose specialized hazards to 23 nearby nuclear power stations. Potential threats to stations from aircraft 24 result from the aircraft itself as a missile and from the secondary effects of 25 a crash, e.g., fire. 26 The acceptability of a site depends on establishing that (1) an accident 27 at a nea'rby industrial, military, or transportation facility will not result 28 in radiological consequences that exceeed the dose guideline in the proposed 29 Section 50.34, or (2) the accident poses no undue risk because it is 30 sufficiently unlikely to occur (less than about 10" per year), or (3) the 31 nuclear power station can be designed so its safety will not be affected by 32 the accident. I 33 Potentially hazardous facilities and activities within 5 miles (8 km) of 34 a proposed site, and major airports within 10 miles (16 km) of a proposed 35 site, should be identified. If a preliminary evaluation of potential 36 accidents at these facilities indicates that the potential hazards from shock 37 waves and missiles approach or exceed those of the design basis tornado of the 14
l
's I region or if potential hazards exist such as flammable vapor clouds, toxic
\ 2 chemicals. or incendiary fragments, the suitability of the site should be 3 determined by a detailed evaluation of the degree of risk imposed by the 4 potential hazard. '
5 The identification of design basis events resulting from the presence of 6 hazardous materials or activities in the vicinity of a nuclear power station 7 is acceptable if the design basis events include each postulated type of 8 accident for which a realistic estimate of the probability of occurrence of 9 potential radiation exposures in excess of the dose value in the proposed 10 Section 50.34(a)(1) exceeds approximately 10" per year. Because of the 11 difficulty of assigning precise numerical values to the probability of 12 occurrence of the types of potential hazards generally considered in 13 determining the acceptability of sites for nuclear stations, judgment must be 14 used as to the acceptability of the overall risk presented by an event. 15 In view of the low probability events under consideration, the 16 probability of occurrence of the initiating events leading to potential 17 radiological consequences in excess of the dose guideline in the proposed Section 50.34(a)(1) should be based on assumptions that are as realistic as is 0 18 19 practicable. In addition, because of the low probability events under 20 consideration, valid statistical data are often not available to permit 21 accurate quantitative calculation of probabilities. Accordingly, a 22 conservative calculation showing that the probability of occurrence of 23 potential radiation exposure in excess of the guideline proposed in Section 24 50.34(a)(1) is approximately 10-' per year is acceptable if, when combined 25 with reasonable qualitative arguments, the realistic probability can be shown 26 to be lower. 27 The effects of design basis events have been appropriately considered ;f 28 analyses of the effects of those accidents on the safety-related features of 29 the proposed nuclear power station have been performed and appropriate 30 measures (e.g., hardening, fire protection) to mitigate the consequences of 31 such events have been taken. 32 The studies described in Section 2.2 of the Standard Review Plan, 33 NUREG-0800, " Standard Review Plan for the Review of Safety Analysis Reports 34 for Nuclear Power Plants,"' should be made to evaluate in detail the suitability of a site in regard to potential accidents involving hazardous 4 35 36 materials and activities at nearby industrial, military, and transportation 15
I facilities. Section 2.2.3 of NUREG-0800 describes evaluation procedures and i 2 criteria for potential accidents in the site vicinity. 3 Regulatory Guide 1.78, " Assumptions for Evaluating the Habitability of a 4 Nuclear Power Plant Control Room During a Postulated Hazardous Chemical 5 Release, describes assumptions acceptable to the NRC staff for use in 6 assessing the habitability of the control room during and after a postulated , 7 external release of hazardous chemicals and describes criteria that are 8 generally acceptable to the staff for the protection of the control room 9 operators. ; 10 Regulatory Guide 1.91, " Evaluations of Explosions Postulated To Occur on 11 Transportation Routes Near Nuclear Power Plants,"" describes a method 12 acceptable to the NRC staff for determining distances from a plant to a 13 railway, highway, or navigable waterway beyond which any explosion that might 14 occur on these routes is not likely to have an adverse effect on plant 15 operation or prevent a safe shutdown. 16 Section 3.5.1.6 of the Standard Review Plan (NUREG-0800) describes 17 review procedures regarding potential aircraft hazards. 18 EC0 LOGICAL SYSTEMS AND BIOTA 19 Areas of great importance to the local aquatic ecosystem may present 20 major difficulties in assessing potential impacts on populations of important 21 species or ecological systems. Such areas include those used for breeding 22 (e.g., nesting and spawning), wintering, and feeding, as well as areas where 23 there may be seasonally high concentrations of individuals of important l 16
i o
}
. l l
1 species.' Where the ecological sensitivity of a site under consideration 2 cannot be established from existing information, more detailed studies, as 3 discussed in Regulatory Guide 4.2, may be necessary. Impacts of station 4 construction and operation on the biota and ecological systems may be miti-5 gated by design and operational practices if justifiable relative to costs and 6 benefits. In general, the important considerations in the balancing of costs 7 and benefits are (a) the uniqueness of a habitat or ecological system within 8 the region under consideration and (b) the amount of habitat or ecological 9 system that would be destroyed or disrupted relative to the total amount of 10 the habitat or ecological system present in the region or the vulnerability of 11 the reproductive capacity of important species' populations to the effects of 12 construction and operation of the plant and ancillary facilities. 13 The alteration of one or more of the existing environmental conditions : 14 may render a habitat unsuitable as a breeding or nursery area. In some cases, 15 organisms use identical breeding and nursery areas each year; if the charac-16 teristics of the areas are changed, breeding success may be substantially 17 reduced or enhanced. Destruction of part or all of a breeding or nursery area 18 may cause population shifts that result in increased competition for the 19 remaining suitable areas. Such population shifts cannot compensate for the 20 reduced size of the breeding or nursery areas if the remaining suitable area 21 is already occupied by the species. Some species will desert a breeding area 22 'A species, whether animal or plant, is important (for the purpose of thi; 23 guide) if a specific causal link can be identified between the nuclear power 24 station and the species and if one or more of the following criteria applies: 25 (1) If the species is commercially or recreationally valuable, 26 (2) If the species is endangered or threatened, 27 (3) If the species affects the well-being of some important species 28 within criteria (1) or (2) or if it is critical to the structure and function 29 of a valuable ecological system or is a biological indicator of radionuclides 30 in the environment. ,
- 31 Endangered and threatened species are defined by the Endangered Species 32 Act of 1973 (16 U.S.C.1531 et seq.) as amended, as follows
- "The term 33 ' endangered species' means any species which is in danger of extinction 34 throughout all or a significant portion of its range other than a species of 35 the Class Insecta determined by the Secretary to constitute a pest whose 36 protection under the provisions of this Act would present an overwhelming and 37 overriding risk to man.' 'The term ' threatened species' means any species which is likely to become an endangered species within the foreseeable future 43839 40 throughout all or a significant portion of its range." Lists of endangered and threatened species are published periodically in the Federal Reaister by 41 the Secretary of the Interior.
17
~ _ _ _ _ _ _ _ - - _ - - _ _ - - _ - _
l 1 because of man's activities in the proximity to the area, even in the absence l 2 of physical disturbance of the actual breeding area. i 3 Of special concern relative to site selection are those unique or 4 especially rich feeding areas that might be destroyed, degraded, or made ! 5 inaccessible to important species by station construction or operation. Eval- ! 6 untion of feeding areas in relation to potential construction or operation l 7 impacts includes the following considerations: size of the feeding area ; 8 onsite in relation to the total feeding area offsite, food density, time of i l 9 use, location in relation to other habitats, topography relative to access ! l 10 routes, and other factors (including man's activities). Site modification oay i ! 11 reduce the quality of feeding areas by destruction of a portion of the food l 12 base, destruction of cover, or both.'
- 13. Construction and operation of nuclear power stations can create barriers )
14 to migration, occurring mainly in the aquatic environment. Narrow zones of ; 15 passage for migratory animals in some rivers and estuaries may be restricted ) 16 or blocked by station operation. Partial or complete blockage of a zone of 17 passage may result from the discharge of heat or chemicals to receiving water , 18 bodies or the construction and placement of power station structures in the 19 water body. Stron -swimming aquatic animals often avoid waters of adverse 20 quality, but larval and immature forms are usually moved and dispersed by l 21 water currents. It is therefore important in site selection that the routes l 22 and times of movement of the immature stages be considered in relation to l 23 potential effects. j 24 A detailed assessment of potential impact on the species population ' 25 would be required for sites where placement of intake or discharge structures 26 would markedly disrupt normal current patterns in migration paths of important 27 species. The potentials for impingement of organisms on cooling water intake 28 structures and entrainment of organisms through the cooling system are deter-29 mined by a number of variables, including site characteristics, intake struc-30 ture design, and placement of the structures at the site. 31 Site characteristics should be considered relative to design and 32 placement of cooling system features and the potential of the cooling system 33 to hold fish in an area longer than the normal period of migration or to 34 entrap resident populations in areas where they would be adversely affected, 35 either directly or indirectly, by limited food supply or adverse temperatures. I 36 Canals or areas where cooling waters are discharged may induce fish to remain 37 in an unnaturally warmed habitat. The cessation of station operation during 1 I 18
I winter can be lethal to these fish because of an abrupt drop in water 2 temperature. 3 LAND USE AND AESTHETICS 4 Many impacts on land use at the site and in the site neighborhood 5 arising from construction and operation of the plant, transmission lines, and 6 transportation corridors can be mitigated by appropriate designs and 7 practices. Aesthetic impacts can be reduced by selecting sites where existing l 8 topography and forests can be utilized for screening station structures from l 9 nearby scenic, historical, or recreational resources. Restoration of natural j 10 vegetation, creative landscaping, and the integration of structures with ) 11 the environment can mitigate adverse visual impacts. 12 Preconstruction archeological excavations can usually reduce losses. 13 Short-term salvage archeology may not be sufficient if extensive or valuable 14 archeological sites are found on the potential site for a nuclear station. 15 For areas of archeological concern, the Cf.ief Archeologist of the National 16 Park Service is an information source, as are the State Archeologist and the i 17 State Liaison Officer responsible for the National Historic Preservation Act 18 activities for a particular State. 19 Proposed alternative land use may render a site unsuitable for a nuclear 20 power station. For example, lands specified by a community (1) as planned for 21 other uses or (2) as restricted to compatible uses vis-a-vis other lands may i 22 be unsuitable. Therefore, official land use plans developed by governments at 23 . any level and by regional agencies should be consulted for possible conflicts l 24 with power station siting. A list of Federal ager.cies that have jurisdiction 25 or expertise in land use planning, regulation, or management has been pub-26 lished by the Council on Environmental Quality.** l 27 Another class of impacts involves the preempting of existing land use at 28 the site itself. For example, nuclear power station siting in areas uniquely i i 29 Station protection requirements for nuclear safeguards may influence landscape 30 design and clearing of vegetation. 31 "See U.S. Council on Environmental Quality, " National Environmental Policy Act
! 32 (NEPA) Implementation Procedures; Appendixes I, II, and III," 49 FR 49750, 33 December 21, 1984.
19 l 1
I suited for growing specialty crops may be considered a type of land conversion 2 involving unacceptable economic dislocation. 3 Sites adjacent to lands devoted to public use may be considered 4 unsuitable. In particular, the use of some sites or transmission lines or 5 transportation corridors close to special areas administered by Federal, 6 State, or local agencies for scenic or recreational use may cause unacceptable 7 impacts regardless of design parameters. Such cases are most apt to arise in 8 areas adjacent to natural-resource oriented areas (e.g., Yellowstone National 9 Park) as opposed to recreation-oriented areas (e.g., Lake Mead National 10 Recreation Area). Some historical and archeological sites may also fall into 11 this category. The acceptability of sites near special areas of public use 12 should be determined by consulting cognizant government agencies. 13 The following Federal agencies should be consulted for the special areas 14 listed: 15
- National Park Service (U.S. Department of the Interior) 16 National Parks; International Parks; National Memorial Parks; National 17 Battlefields, Battlefield Parks, and Battlefield Sites; National 18 Military Parks; Historic Areas and National Historic Sites; National 19 Capital Parks; National Monuments and Cemeteries; National Seashores and 20 Lakeshores; National Rivers and Scenic Riverways; National Recreation 21 Areas; National Scenic Trails and Scientific Reserves; National Parkways 22
- National Park Service Preservation Program 23 National Landmarks Program; Historic American Buildings Survey; National 24 Register of Historic Places; National Historical Landmarks Program; 25 National Park Service Archeological Program 26
- Bureau of Sport Fisheries and Wildlife (U.S. Department of Interior) 27 National Wildlife Refuges 20
I 1
- Forest Service (U.S. Department of Agriculture) 2 National Forest Wilderness, Primitive Areas, National Forests.
3 Individual States and local governments administer parks, recreation 4 areas, and other public use and benefit areas. Information on these areas 5 should be obtained from cognizant State agencies such as State departments of 6 natural resources. The Advisory Council on Historic Preservation or the 7 appropriate State historic preservation officer should be contacted for 8 information on historic areas. 9 - It should be recognized that some areas may be unsuitable for siting 10 because of public interest in future dedication to public scenic, 11 recreational, or cultural use. Relatively rare land types such as sand dunes 12 and wetlands are examples. However, the acceptability of sites for nuclear 13 power stations at some future time in these areas will depend on the. existing 14 impacts from industrial, commercial, and other developments. 15 SOCI0 ECONOMICS 16 Social and economic issues are important determinants of siting policy. 17 It is difficult both to assess the nature of the impacts involved and to 18 determine value schemes for predicting the level or the acceptability of 19 potential impacts. 20 The siting, construction, and operation of a nuclear power station may 21 have significant impacts on the socioeconomic structure of a comunity and may 22 place severe stresses on the local labor supply, transportation facilities, 23 and comunity services in general. There may be changes in the tax basis and 24 in comunity expenditures, and problems may occur in determining equitable 25 levels of compensation for persons relocated as a result of the station sit-l 26 ing. It is usually possible to resolve such difficulties by proper coordina-27 tion with impacted comunities; however, some impacts may be locally unaccept-28 able and too costly to avoid by any reasonable program for their mitigation. 29 Evaluation of the suitability of a site should therefore include consideration 30 of purpose and probable adequacy of socioeconomic impact mitigation plans for 21 l
I such economic impacts on any community where local acceptance problems can be 2 reasonably foreseen. 3 l Certain communities in the neighborhood of a site may be subject to 4 unusual impacts that would be excessively costly to mitigate. Among such 5 communities are towns that possess notably distinctive cultural character, 6 1.e., towns that have preserved or restored numerous places of historic 7 interest, have specialized in an unusual industry or avocational activity, or 8 have otherwise markedly distinguished themselves from other communities. 9
. Siting decisions should reflect fair treatment and meaningful 10- involvement of all people, regardless of race, ethnicity, culture, income or 11 educational level to assure equitable consideration and to minimize 12 disproportionate effects on minority and low-income populations.**
13 NOISE 14 Noise levels at nuclear stations occur during both the construction and 15 operation phases and could have unacceptable impacts. Cooling towers, tur-16 bines, and transformers contribute to the noise levels during station 17 operation. ' l 18 C. REGULATORY POSITION 19 1. GE0 LOGY / SEISM 0 LOGY 20 Preferred sites are those with a minimal likelihood of surface or near-21 surface deformation and a minimal likelihood of earthquakes on faults in the 22 site vicinity (within a radius of 8 km (5 miles)). Because of the 23 uncertainties and difficulties in mitigating the effects of permanent ground 24 displacement phenomena such as surface faulting or folding, fault creep, 25 subsidence or collapse, the NRC staff considers it prudent to select an 26 alternative site when the potential for permanent ground displacement exists 27 at the site. 28 The NRC committed to carry out the measures set forth in Executive Order 12898, 29 " Federal Actions to Address Environmental Justice in Minority Populations and A 30 Low-Income Populations" (59 FR 7629), to consider the effects of its actions on 4 31 minority and low-income communities. 22 l
4 1 Sites located near geologic structures, for which at the time of 2 application the data base is inadequate to determine their potential for 3 causing surface deformation, are likely to be subject to a longer licensing 4 process in view of the need for extensive and detailed geologic and seismic 5 investigations of the site and surrounding region and for the rigorous ! 6 analyses of the site-plant combination. 7 Sites with competent bedrock generally have suitable foundation j 8 conditions. In regions with few or no such sites, it is prudent to select
- 9 sites with competent and stable solid soils, such as dense sands and glacial
, 10 tills. Other materials may also provide satisfactory foundat.on conditions, ! 11 but a detailed geologic and geotechnical investigation would be required to ; 12 determine static and dynamic engineering properties of the material underlying l 13 the site in accordance with the proposed Section 100.23 to 10 CFR Part 100. 14 2. ATMOSPHERIC EXTREMES AND DISPERSION 15 As noted in the Discussion Section of this guide, site atmospheric ; 16 conditions are site suitability characteristics, principally with respect to 17 the calculation of radiation doses resulting from the release of fission l 18 products as a consequence of a postulated accident. Accordingly, each 19 applicant for site approval must collect meteorological information for at 20 least one year that is representative of the site conditions, including wind . 21 speed, wind direction, precipitation, and atmospheric stability. ] 22 Nonradiological atmospheric considerations such as local fogging and 23 icing, cooling tower drift, cooling tower plume lengths, and plume 24 interactions between cooling tower plumes, as well as plumes from nearby 25 industrial facilities, should be considered in evaluating the suitability of - \ 26 potential sites. 27 3. EXCLUSION AREA AND LOW POPULATION ZONE 28 An applicant for a reactor license is required by 10 CFR Part 100 to 29 designate an exclusion area and to have authority to determine all activities 30 within that area, including removal of personnel and property. Transportation 31 corridors such as highways, railroads, and waterways are permitted to traverse 32 the exclusion area provided (1) these are not so close to the facility as to 33 interfere with normal operation of the facility and (2) appropriate and 23
e 1 effective arrangements are made to control traffic on the highway, railroad, 2 or waterway in the case of emergency to protect the public health and safety. 3 The exclusion area must be of such a size that an individual assumed to 4 be located at any point on its boundary would not receive a radiation dose in 5 excess of 25 rem total effective dose equivalent (TEDE) over any two-hour 6 period following a postulated accidental fission product release into the 7 containment. 8 An applicant is also required by 10 CFR Part 100 to designate an area 9 immediately beyond the exclusion area as a low population zone (LPZ). The size l 10 of the LPZ must be such that the distance to the boundary of the nearest j 11 densely' populated center containing more than about 25,000 residents must be 12 at least one and one-third times the distance from the reactor to the outer 13 boundary of the LPZ. The boundary of the population center should be I 14 determined upon consideration of population distribution, not political 15 boundaries. 16 The proposed Section 50.34 would require that the LPZ be of such a size l 17 that an individual located on its outer radius for the course of the 18 postulated accident (assumed to be 30 days) would not receive a radiation dose 19 in excess 0f~25 rem TEDE. l l 20 4. POPULATION CONSIDERATIONS l 21 The proposed paragraph 100.21(h) states that, " Reactor sites should be 22 located away from very densely populated centers. Areas of low population 23 density are, generally, preferred. However, in determining the acceptability 24 of a particular site located away from a very densely populated center but not 25 in an area of low density, consideration will be given to safety, 26 - environ:nental, economic, or other factors, which may result in the site being 27 found acceptable." 28 Locating reactors away from densely populated centers is part of the , 29 NRC's defense-in-depth philosophy and facilitates emergency planning and ' 30 preparedness as well as reducing potential doses and property damage in the 31 event of a severe accident. Numerical values in this guide are generally 32 consistent with past NRC practice and reflect consideration of severe 33 accidents, as well as the demographic and geographic conditions characteristic 34 of the United States, j 24
1 A reactor preferably should be located such that, at the time of initial 2 site approval and within about 5 years thereafter, the population density, 3 including weighted transient population, averaged over any radial distance out 4 to 20 miles (cumulative population at a distance divided by the circular area 5 at that distance), does not Exceed 500 persons per square mile. A reactor 6 should not be located at a site whose population density is well in excess of 7 the above value. 8 If the population density of the proposed site exceeds, but is not well 9 in excess of the above preferred value, the analysis of alternative sites 10 shou.ld pay particular attention to alternative sites having lower population 11 density. However, consideration will be given to other factors such as 12 safety, environmental, or economic considerations, which may result in the 13 site with the higher population density being found acceptable. Examples of 14 such factors include, but ari not limited to, the higher population density 15 site having superior seismic characteristics, better rail or highway access, 16 shorter transmission line requirements, or less environmental impact upon 17 undeveloped areas, wetlands, or endangered species. 18 The transient population should be included for those sites where a 19 significant number of people (other than those just passing through the area) 20 work, reside part-time, or engage in recreational activities and are not 21 permanent residents of the area. The transient population should be taken 22 into account for site evaluation purposes by weighting the transient 23 population according to the fraction of time the transients are in the area. 24 Projected changes in population within about 5 years after initial site 1 I 25 approval should be evaluated for the proposed site and any alternative sites 26 considered. Population growth in the site vicinity after initial site 27 approval is normal and expected and will be periodically factored into the 28 emergency plan for the site, but population increases after initial site 29 approval will not be a factor in license renewal or, by itself, used to impose 30 other license conditions or restrictions on an operating plant. 31 5. EMERGENCY PLANNING 32 The proposed Section 100.21(g) states that " Site characteristics must be 33 such that adequate plans to take protective actions for members of the public 34 in the event of emergency can be developed." 25
. 1 i
i 1 An examination and evaluation of the site and its vicinity, including g 2 the population distribution and transportation routes, should be conducted to T 3 determine whether there are any characteristics that would prevent taking , l 4 protective actions to protect the public in the event of emergency. 5 Special population groups, such as those in hospitals, prisons, or other 6 facilities that could require special needs during an emergency, should be ! 7 identified. ' I B Physical characteristics of the proposed site that could pose a 9 significant impediment to taking protective measures, such as egress l 10 limitations from the area surrounding the site, should be identified.
- 11 An evacuation time estimate (ETE) should be performed to estimate the l
12 time periods that would be required to evacuate various sectors of the plume 13 exposure emergency planning zone (EPZ), including the entire EPZ. The ETE is ( 14 an emergency planning tool that assesses, in an organized and systematic ' 15 fashion, the feasibility of taking protective measures for the population in 16 the surrounding area. Information on performing an ETE analysis is given in 17 Appendix 4 to NUREG-0654/ FEMA-REP-1, Revision 1, " Criteria for Preparation and i 18 Evaluation of Radiological Emergency Response Plans and Preparedness in g 19 Support of Nuclear Power Plants" (November 1980).* The value of the ETE W l 20 analysis is in the methodology required to perform the analysis rather than in 21 the calculated ETE times. While lower ETEs may reflect favorable site 22 characteristics from an emergency planning standpoint, there is no minimum 23 required evacuation time in the regulations that an applicant has to meet. 24 6. SECURITY PLANS 25 The. proposed Section 100.21(f) states " Site characteristics must be such 26 that adequate security plans and measures can be developed." Also,10 CFR 27 Part 73 describes physical protection requirements for nuclear power plants as 28 well as special nuclear materials. 29 Generally, a distance of about 110 meters (360 feet) to any vital 30 structure or vital equipment would provide sufficient space to satisfy 31 security measures of 10 CFR 73.55 (e.g., protected area barriers, detection 32 equipment, isolation zones, vehicle barriers). If the distance to a vital 33 structure or vital equipment is less than about 110 meters (360 feet), special l 34 measures or analyses may be needed to show that adequate security plans can be ! 35 developed. 26
l 1
\ l 7. HYOROLOGY 2 7.1 Floodino 3 To evaluate sites located in river valleys, on flood plains, or along 4' coastlines where there is a potential for flooding, the site suitability 5 studies described in Regulator,y 1.59, " Design Basis Floods for Nuclear Power '
6 Plants,"' should be made. 7 7.2 Water Availability 8 A highly dependable system of water supply sources must be shown to be 9 available under postulated occurrences of natural and site-related accidental 10 phenomena or combinations of such phenomena as discussed in Regulatory Guide 11 1.59. 12 To evaluate the suitability of sites, there should be reasonable 9 13 14 assurance that permits for consumptive use of water'in the quantities needed for a nuclear power plant of the stated approximate capacity and type of 15 cooling system can be obtained by the applicant from the appropriate State, 16 local, or regional bodies. 17 7.3 Water Quality 18 The potential impacts of nuclear power stations on water quality are 19 likely to be acceptable if effluent limitations, water quality criteria for 20 receiving waters, and other requirements promulgated pursuant to the Federal 21 Water Pollution Control Act are applicable and satisfied. 22 The criteria in 10 CFR Parts 20 and 50 will be used by the NRC staff for 23 determining permissible concentrations of radioactive niaterials discharged to 24 surface water or to ground water.88 6 25 26 27
" Appendix I to 10 CFR Part 50 provides numerical guidance for design objectives and technical specification requirements for limiting conditions of operation for light-water-cooled nuclear power stations.
27
1 7.4 Fission Product Retention and Transport 2 To be able to assess fission product retention and transportation via 3 groundwater, the following information should be determined for the site: 4 e- Soil, sediment, and rock characteristics (e.g., volcanic ash, fractured limestone, etc.), 5 e
~
6 Absorption and retention coefficients for radioactive materials, 7 e' Ground-water velocity, and
'8 e Distance to nearest body of surface water. l 1
9 This information should be used in the environmental report required in 10 CFR
- 10 Part 51 and compared to the hydrological information used in the PRA or other 11 analyses for a certified-plant design (if such a design is to be located at 12 the site) or used in the site-specific PRA for a custom plant located at the i 13 site.
14 Aquifers that are or may be used by large populations for domestic, 15 municipal, industrial, or irrigation water supplies provide potential pathways 16 for the transport of radioactive material to man in the event of an accident. 17 To evaluate the suitability of proposed sites located over such aquifers, L 18 detailed studies of factors identified in Section 2.4.13 of Regulatory Guide 19 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear 1 20 Power Plants,"' should be completed. l 21 8. INDUSTRIAL. MILITARY. AND TRANSPORTATION FACILITIES 22 'The proposed.Section 100.21(e) states " Potential hazards associated with 23 nearby transportation routes, industrial and military facilities must be 24 evaluated and site parameters established such that potential hazards from 25 such routes and facilities will pose no undue risk to the type of facility 26 proposed to be located at the site." 27 The acceptability of a site would depend on establishing that (1) an i j -28 accident at a nearby industrial, military, or transportation facility would i ! 28 i
6
- j. .
1 not result in radiological consequences that exceed the dose guideline in the 2 proposed Section 50.34, or (2) the accident poses no undue risk because it is 3 sufficiently unlikely to occur (less than about 10" per year), or (3) the 4 nuclear power station can be designed so its safety will not be affected by 5 the accident.
; 6 Potentially hazardous facilities and activities within 8 km (5 mi) of a 7 proposed site, and major airports within 16 km (10 mi) of a proposed site, 8 should be identified. If a preliminary evaluation of potential accidents at 9 these facilities indicates that the potential hazards from shock waves and 10 missiles approach or exceed those of the design basis tornado for the region 11 or,there are potential hazards such as flammable vapor clouds, toxic chemi-12 cals, or incendiary fragments, the suitability of the site should be 13 determined by detailed evaluation of the degree of risk imposed by the 14 potential hazard. The design basis tornado is described in Regulatory Guide 15 1.76, " Design Basis Tornado for Nuclear Power Plants."'
16 The identification of design basis events resulting from the presence of l 17 hazardous materials or activities in the vicinity of a nuclear power station 18 is acceptable if the design basis events include each postulated type of 19 accident for which a realistic estimate of the probability of occurrence of 20 doses in excess of the guideline proposed in Section 50.34(a)(1) exceeds 21 approximately 10" per year. Because of the difficulty of assigning precise 22 numerical values to the probability of occurrence of the types of potential 23 hazards generally considered in determining the acceptability of sites for 24 nuclear stations, judgment must be used as to the acceptability of the overall 25 risk presented by an event. l 26 In view of the low-probability events under consideration, the 27 probability of occurrence of initiating events leading to potential con-28 sequences in excess of the dose guideline proposed in Section 50.34(a)(1) 29 should be based on assumptions that are as realistic as is practicable. 30 Because of the low-probability events under consideration, valid statistical 31 data are often not available to permit accurate quantitative calculation of 32 probabilities. Accordingly, a conservative calculation showing that the 33 probability of occurrence of potential doses in excess of the guideline 34 proposed in Section 50.34(a)(1) is approximately 10-' per year is acceptable 35 if, when combined with reasonable qualitative arguments, the realistic 4 36 probability can be shown to be lower. 29
1 The effects of design basis events have been appropriately considered if 2 analyses of the effects of those accidents on the safety-related features of a 3 proposed nuclear station have been performed and appropriate measures (e.g., , 4 hardening, fire protection) to mitigate the consequences of such events have 5 been taken. . 1 6 9. EC0 LOGICAL SYSTEMS AND BI0TA 8 The ecological systems and biota at potential sites and their 9 environs should be sufficiently well known to allow reasonably certain 10 predicfions that there would be no unacceptable or unnecessary deleterious 11 impacts on populations of important species or on ecological systems with 12 which they are associated from the construction or operation of a nuclear 13 ' power station at the site. 14 When early site inspections and evaluations indicate that critical or 15 exceptionally complex ecological systems will have to be studied in detail to 16 determine the appropriate plant designs, proposals to use such sites should be 17 deferred unless sites with.less complex characteristics.are not available. 18 It should be determined whether any important species (as defined in the 1 19 Discussion section of this guide under Ecological Systems and Biota) inhabit l 20 or use the proposed site or its environs. If so, the relative abundance and 21 distribution of their populations should be considered. Potential adverse j 22 impacts on important species should be identified and assessed. The relative 23 abundance of individuals of an important species inhabiting a potential site 24 should be compared to available information in the literature concerning the 25 total estimated local population. Any predicted impacts on the species should 26 be evaluat'ed relative to effects on the local population and the total 27 population of the species. The destruction of, or sublethal effects on, a 28_ number of individuals that would not adversely affect the reproductive 29 capacity and vitality of a population or the crop of an economically important 30 harvestable population or recreationally important population should generally 31 be acceptable, except in the case of certain endangered species. If there are 32 endangered or threatened species at a site, the potential effects should be 33' evaluated relative to the impact on the local population and the total 34 estimated population over the entire range of the species as noted in the 35 literature. 30
1 It should be determined whether there are any important ecological l ( 2 systems at a site or in its environs. If so, determination should be made as 3 to whether the ecological systems are especially vulnerable to change or if 4 they contain important species habitats, such as breeding areas (e.g., nesting 5 and spawning areas), nursery, feeding, resting, and wintering areas, or other 6 areas of seasonally high concentrations of individuals of important species. 7 Important considerations in balancing costs and benefits include the 8 uniqueness of a habitat or ecological system within the region under 9 consideration, the amount of the habitat or ecological system destroyed or 10 disrupted relative to the total amount in the region, and the vulnerability of 11 the, reproductive capacity of important species populations to the effects of 12 construction and operation of the station and ancillary facilities. 13 If sites contain, are adjacent to, or may impact on important ecological 14 systems or habitats that are unique, limited in extent, or necessary to the 15 productivity of populations of important species (e.g., wetlands and estuar-16 ies), they cannot be evaluated as to suitability for a nuclear power station 17 until adequate assessments for the reliable prediction of impacts have been 18 completed and the facility design characteristics that would satisfactorily 19 mitigate the potential ecological impacts have been defined. In areas where 20 reliable and sufficient data are not available, the collection and evaluation 21 of appropriate seasonal data may be required. 22 Migrations of important species and migration routes that pass through 23 the site or its environs should be identified. Generally, the most critical 24 migratory routes relative to nuclear power station siting are those of aquatic 25 species in water bodies associated with the cooling systems. Site conditions 26 that should be identified and evaluated in assessing potential impacts on 27 important aquatic migratory species include (1) narrow zones of passage, 28 (2) migration periods that are coincident with maximum ambient temperatures, 29 (3) the potential for major modification of currents by station structures, 30 (4) the potential for increased turbidity during construction, and (5) the 31 potential for entrapment, entrainment, or impingement by or in the cooling 32 water system or for blocking of migration by facility structures or effluents. 33 The potential for blockage of movements of important terrestrial' animal 34 populations caused by the use of the site for a nuclear power station ar.d the 35 availability of alternative routes that would provide for maintenance of the 36 species' breeding population should be assessed. 31
j- . I 1 If justifiable relative to costs and benefits, the potential impacts of . 2 plant construction and operation on the biota and ecological systems can 3 generally be mitigated by adequate engineering design and site planning and by I 4 proper construction and operations when there is adequate information about 5 5 the vulnerability of the important species and ecological systems.
- 6 A sumary of environmental considerations, parameters, and regulatory l 7 positions for use in evaluating sites for nuclear power stations is provided
} 8 in Appendix B to this guide. i 9 10. LAND'USE AND AESTHETICS j 10 L'and use plans adopted by Federal, State, regional, or local i 11 governmental entities should be examined, and any conflict between these plans 12 and use of a potential site should be resolved by consultation with the 13 appropriate governmental entity. 14 For a potential site on land devoted to specialty crop production where 15 changes in land use might result in market dislocations, a detailed 16 ' investigation should be provided to demonstrate that potential impacts have 17 been identified. 18 The potential aesthetic impact of nuclear power stations at sites near 19 natural-resource-oriented public use areas is of concern, and evaluation of 20 such sites is dependent on consideration of specific station design layout. 21 11 1QG.LQiG9t!9t!LCji 22 The NRC staff considers that an evaluation of the suitability of nuclear 23 power station sites near distinctive comunities should demonstrate that the 24 construction and operation of the nuclear station, including transmission and 25 transportation corridors, and potential problems relating to comunity serv- ' 26 ices, such as schools, police and fire protection, water and sewage, and 27 health facilities, will not adversely affect the distinctive character of the 28 comunity nor disproportionately affect minority or low-income populations. A 29 preliminary investigation should be made to address environmental justice 30 considerations and to identify and analyze problems that may arise from the 31 proximity of a distinctive comunity to a proposed site. 32
- m. . - . -
I 1 12. NOISE 2 Noise levels at proposed sites must comply with applicable Federal, 3 State, and local noise regulations. 4 D. IMPLEMENTATION 5 The purpose of this section is to provide guidance to applicants and 6 licensees regarding the NRC staff's plans for using this regulatory guide. 7 This proposed revision has been released to encourage public 8 participation in its development. Except in those cases in which the 9 applicant proposes an acceptable alternative method for complying with the 10 specified portions of the Commission's regulations, the method to be described 11 in the active guide reflecting public comments will be used in the evaluation 12 of applications for construction permits, operating licenses, combined 13 licenses, or design certification submitted after the implementation date to 14 be specified in the active guide. This guide would not be used in the 15 evaluation of an application for an operating license submitted after the 16 implementation date to be specified in the active guide if the construction 17 permit was issued prior to that date. l l l 4 33
ee 1 APPENDIX A 2 SAFETY-RELATED SITE CONSIDERATIONS 3 FOR ASSESSING SITE SUITABILITY 4 FOR NUCLEAR POWER STATIONS 5 This appendix provides a checklist of safety-related site charac-6 teristics, relevant regulations and regulatory guides, and regulatory 7 experience and positions for assessing site suitability for nuclear power 8 statio'ns. l I l l l f
- 4 l
A-1
Considerations Relevant Regulations and Regulatory Guides Regulatory Experience and Position g l 1 A.1 Geology / Seismology 1 2 Geologic and seismic char- Proposed amrendment to 10 3 Where the potential for permanent acteristics of a site, such as CFR Part 100, proposed 4 ground deformation such as j surface faulting, ground Section 100.23, "Geologc 5 faulting, folding, subsidence or ' motion, and foundation condi- and Seismic Siting Factors" 6 collapse exists at a site, the NRC tions (including liquefaction, (59 FR 52255). staff considers it prudent to select 7 subsidence, and landslide t 8 potential), may affect the an attemative site. Regulatory Guide 1.70, 9 safety of a nuclear power Chapter 2 (identifies safety- Sites should be selected in areas 10 station. related site characteristics).' for which an adequate geologic 11
- data base exists or can be Regulatory Guide 1.29 expeditiously developed througi' 12 (discussu plant safety site-specdic investigstens to 13 features which should be identify and characterize potential 14 controlled by engineering gelogical and seismic hazards.
15 design).' Delay in licensing can result from 16 a need for extensive geologic and Draft Regulatory Guide seismic investigations. 17 DG-1032, " Geological, 18 Conservative design of safety-Seismological, and related structures will be required 19 Geophysical Investigations to when geologic, seismic, and 20 Characterize Seismic foundation information is ques-21 Sources."2 tionable, f y , 22 Regulatory Guide 1.132, " Site 23 Sites with competent bedrock investigations for Foundations generally have suitable foundation 24 of Nuclear Power Plants".' conditions. 25 26 If bedrock sites are not available, 27 it is prudent to select sites in 28 areas known to have a low subsi-29 dence and liquefaction potential. 30 invatigations will be required to 31 determine the static and dynamic 32 engineering properties of the 33 material underlying the site as 34 stated in 10 CFR Part 100, Appendix A and the proposed g Section 100.23.
)
37
' Copia are available for inspection or copying for a fu from the NRC Public Document Room at 2120 38 39 L Street NW., Washington, DC: the PDR's mailing address is Mail Stop LL-6, Washington, DC 20555;
"$.ac.e (202)634-3273: fax (2021634-3343. Copies may be purchased at current rates frm the 40 U.S. Government Printing Office, P.O. Box 37082, Washington, DC 20402 9328 (telephone (202)512-41 2249): or from the National Technical Information Service by writing NTIS at 5285 Port Royal Road, 42 Springfield, VA 22161.
43 44
- Requests for single copies of draft guides should be made in wnting to the U.S. Nuclear Regulatory Commission, Washington, DC 20555, Attention: Off*ce of Administration, Distribution and Mail 45 Services Secten. Copia are available for inspection or copying for a fee from the NRC Public 46 ,
47 Document Room at 2120 L Street NW., Washmaton, DC: the PDH's mailing address is Mail Stop LL-6, Wash 6ngton, DC 20555: telephone (202)634-3273: fax (202)634 3343. A-2
l
]
Considerations Relevant Regulations Regulatory Experience r% and Regulatory Guides and Position b I A.2 Atmospheric Dispersion 2 The atmospheric conditions at 10 CFR Part 50, "Dornoste Unfavorable safety-related design 3 a site should provide sufficient Licensing of Production and basis atmospheric dispersson 4 dispersion of radioactive Utilization Facilities." charactonstes can be
; 5 materials rewased during a compensated for by engineered 6 postulated siccident to reduce Regulatory Guide 1.23, safety features. As-t.;..ii y, l the . 7' the radiation expcsures of "Onsite Meteorological regulatory positen on atmosphenc ) ' Programs."'
8 indisiduals ait the exclusion dispersion of radiological effluents l is incorporated into the section 9 area and low population zone 10 boundaries to the values Regulatory Guide 1.145, " Exclusion Area and Low 11 proposed in Section 50.34. " Atmospheric Dispersion Populaten Zone" (see A.3 of this 12 . Models for Potential Accident appendix). 13 Consequence Assessments at 14 Nuclear Power Plants."' 15 Regulatory Guide 1.3 16 " Assumptions Used for 17 Evaluating the Potential 18 Radiological Consequences of 19 a loss of Coolant Accident for 20 Boiling Water Reactors."' ) l 21- Regulatory Guide 1.4, j 22 "Assumptbns Used for 23 Evaluating the Potential ' 24 Radiological Consequences of 25 a Loss of Coolant Accident for 26 Pressurized Water Reaciors."' 27 Regulatory Guide 1.5, 28
- Assumptions Used for Eval.
29 uating the Potential 30 Radiological Consequences of 31 a Steam Line Break Accident 32 for Boiling Water Reactors."' 33 Regulatory Guide 1.25, 34 " Assumptions Used for 35 Evaluating the Potential i 36 Radiological Consequences of 37 a Fuel Handling Accident in 38 the Fuel Handling and Storage 39 Facility for Boiling and 40 Pressurized Water Reactors."' P G A-3
Considerations Relevant Regulations Regulatory Experience and Regulatory Guides and Position ' 1 A.3 Exclusion Area and 2 Low Population Zone 3 in the event of a postulated 4 10 CFR Part 100, "Rescun M Based on the assumptions in accident at a nuclear power Criteria,' requires the following: 5 station, radiological consequences Regulatory Guides 1.3 and 1.4, the 1 6 required distances to the exclusion for individual members of the
- An " exclusion area' 7 public outside the station must be area boundary and the outer boundary surrounding ti,e reactor in which acceptably low. of the LPZ will depend upon plant the reactor licensee has the design aspects such as the reactor 10 authority to determine all power level, allowable containment 11 activities, including exclusion or rernovel of personnel and leak rate, and those engineered safety 12 features incorporated into the design,.
13 property, and a "iow population as weil as the atmospheric dispersion 14 ' zone" (LPZ) which immediately characteris6cs of the site. surrounds the exclusion area. 15 16
- 10 CFR Part 50, ' Domestic 17 Ucensing of Production and 18 Utilization Facilities." requires 19 that at any point on the 20 exclusion area boundary and on
- 2) the outer boundary of the LPZ 22 the exposure of an individual to 23 a postulated release of fission 24 products (as a consequence of 25 an accident) be less than 25 26 rem total effective dose 27 equivalent, for specified time ..
periods. 28 29 e Regulatory Guides 1.3,1.4, 30 1.5, and 1.25 give calculational 31 methods (see A.2 of this appendix.) A-4
, O r I Considerations Relevant Regulations Regulatory Experience and Regulatory Guides and Position j 3 A.4 Population Considerations 10 CFR Part 100, " Reactor Site A reactor should preterably be located 2 Locating reactors away from Criteria," requires the following: such that, at the time of initial site 3 densely populated centers is part 4 of the NRC's defense-in-depth approval and within sbout 5 years
- An " exclusion area
- thereafter, the population denssty, 5 philosophy and facilitates i emergency planning and surrounding the reactor in which including weighted transient 6
the reactor licensee has the population, averaged over any radial 7 preparedness as well as reducing 8 potential doses and property authority to determine all distance out to 20 miles (cumulative damage in the event of a severe activities, including exclusion or population at a distance divided by 9 removal of personnel and the area at that distance), does not 10 accident. 11 property, and a " low population exceed 500 persons per square mile.
~12 zone * (LPZ), which immediately A reactor should not be located at a 13 -
surrounds the exclusion area. site whose population density is well in excess of the above value. 14 e The distance to the nearest 15 densely populated center if the population density of the 16 containing more than about proposed site exceeds, but is not well 17 25,000 residents must be at in excess of, the preferred value, the
'8 least one and one-third times analysis of attemative sites should L' the distance from the reactor to pay particular attention to alternative J the outer boundary of the LPZ. sites having lower population density.
Consideration will be given to other 21 e Reactor sites should be located factors, such as safety, U away from very densely environmental, or r.conomic, which 9 23 24 25 populated centers. Areas of low population density are, generally, preferred. However, in determining the acceptability of may result in the site with higher population density being found acceptable. 26 27 a particutar site located away Transient population should be 28 from a very densely populated included for those sites where a 29 center but not in an area of low significant number of people (other 30 density, consideration will be than those just passing through the 31 given to safety, environmental, area) work, reside part-time, or 32 . economic, or other factors, engage in recreational activities, and 1 33 which may result in the site are not permanent residents of the l 34 being found acceptable. area. The transient population should 35 be taken into account by weighing the I
, 36 transient population according to the
- 37 fraction of time the transients are in 38 the area.
l l 39 4 1 I i 1 A-5 l
Considerations Relevant Regulations i Regulatory Experience and Regulatory Guides and Position l } A.5 Ernergency Planning I 2 To ensure that adequate 3 10 CFR Part 100, *Ikactor Site protective measures can be taken criteria, requires that: An exarmnation and evaluanon of the 4 to protect members of the public
- site should be conducted to deterrmne 5 in the event of an omwgency, the whether there are any charactenstes i 6
- Site charactenstes must be that would prevent talung protective i charactwistics of tne site should such that adequate plans to take 7 not preclude development of such actions to protect the public in the 8 protectrve schons for membes event of emergency.
plans. 9 of the public in the event of i emwgem:y can be developed. Physical charactenstics of the proposed site that could pose a 10 significant impedunent to talung
}} 10 CFR Part 50,
- Domestic Ucensing of Production and protect ve actens, such as egress 12 limitations from the area surrounding 13 - Utilization Facilities," requires: the site, should be identified.
14 15
- Reasonable assurance that Special pop'dation groups, such as 16 adequate protection can and will those in hospitals, prisons, or other 17 be taken in the event of a facilities that could require s radiological emergency. needs during an emergency,pecialshould be 18 identified.
19
- Emergency planning zones 20 (EPZ) consisting of the plume exposure pathway EPZ with an An evacuation time estwnate IETE) 21 should be performed to estimate the 22 area about 16 km (10 mi) in time penods that would be required to 23 radius, and the ingestion evacuate various sectors of the plume 24 pathway EPZ with an area about exposure emergency planning zone 80 km (50 mi) in radius. (EPZ), including the entire EPZ. The ',
25 ETE analysis is an emergency planrung 26 NUREG-0654/ FEMA-fMP-1, tool that assesses, in an organized 27 Rev.1, ' Criteria for Preparation and systematic fashion, the feasibility 28 and Evaluation of Radiological of talung protective measures for the 29 Emwgency Response Plans and populaDon in the surroundmg area. - 30 Preparedneu in Support of While loww ETEs may reflect 31 Nuclear Power Plants' (November 1980),' provides favorable site charactensacs from an 32 asnergency plannmg standpoint, there 33 guidan.e on performing an ETE. is no mnumum required evacuation time an applicant must meet. 4 A-6
f Considerations Relevant Regulations Regulatory Experience ' and Regulatory Guides and Position 1 A.6 Security Plans To prevent plant damage, and The proposed Section 100.21(f) Generally, a distance of about 110 2 possible radiological states: meters to any vital structure or vital 3 consequences to the public as a equipment would provide space
- 4 result of acts of sabotage, the Site characteristics must be sufficient to satisfy security measures i 5 such that adequate secunty specified in 10 CFR Part 73.55 (e.g.,
I 6 characteristics of the site should not preclude development of plans and measures can be protected area barriers, detection
; 7 adequate security plans. developed. equipment, isolation zones, vehicle 8
barriers). 9 Also,10 CFR Part 73, " Physical if the distance to a vital structure or 10 Protection of Plants and vital equipment is less than about 110 Materials," prescribes meters, special measures or analyses 11 - 12 requirements for establishment may be required to show that and maintenance of a physical adequate security plans can be 13 14 protection system for the developed. 15 protection of special nuclear 16 materials at fixed sites and of
' 17 plants in which special nuclear 18 materialis used.
19
/
t l l l i A-7
Considerations Relevant Regulations ! Regulatory Experience end Regulatory Guides and Position 1 A.7 Hydrology 2 A.7.1 Flooding 3 Precipitation, wind, or 4 The proposed Section 100.23, seismically induced flooding To evaluate sites located in river 5 " Geologic and Seismic Siting valleys, on flood plains, or along (e.g., resulting from dam Factors." 6 failure, from river blockage or coastlines where there is a 7 diversion, or from distantly potential for flooding, the studies Regulatory Guide 1.59, described in Regulatory Guide 8 and locally generated sea
- Design Basis Floods for 9 waves) can affect the safety 1.59 should be made.
10 Nuclear Power Plants."' of a nuclear power station. 11
' Regulatory Guide 1.70, i 12 13 " Standard Format and Content 14 of Safety Analysis Reports for 15 Nuclear Power Plants" (Section 2.4).'
16 17 10 CFR Part 50, Appendix A,
" General Design Criteria for 18 Nuclear Power Plants;"
19 Criterion 2, " Design Bases for 20 Protection Against Natural 21 Phenomena." 22 A.7.2 Wste Availability 23 A safett-rela ed water supply The proposed Section 100.23, A highly dependatale system of 24 is required fcr normal or
" Geologic and Seismic Siting water supply sources should be 25 emergency shutdown and Factors.* shown to be available under 26 cooldown.
27 postulated occurrences of natural Regulatory Guide 1.59, phenomena and site-related 28 " Design Basis Floods for 29 accidental phenomena or Nuclear Power Plants."' combinations of such phenomena 30 as discussed in Regulatory Guide Regulatory Guide 1.27, 1.59. 31
" Ultimate Heat Sink for 32 Nuclear Power Plants." To evaluate the suitability of a 33 34 site, there must a reasonable 35 assurance that permits for water 36 use and for water consumption in 37 the quantities needed for a 38 nuclear power plant of the stated 39 approximate capacity and type of 40 cooling system can be obta'med 41 by the applicant from the ap-42 propriate State, local, or regional bodies.
A-8
. \
Relevant Regulations Regulatory Experience Considerations
' and Regulatory Guides and Position -
1 A.7.3 Water Quality f I 10 CFR Part 20, " Standards The criteria provided in 10 CFR i 2 Contamination of ground for Protection Against Parts 20 and 50 will be used by J 3 water and surface water by
' Radiation." the NRC staff for determining i 4 radioactive materials permissible concentrations of l j 5 discharged from nuclear 10 CFR Part 50, " Licensing of radionuclides discharged to 6 stations could cause public ' Production and Utilization surface water and ground water, ,
7 health hazards. 8 Facilities." l i t I V . A-9
.-- - . ~ - - - - - . -
i,
- h i
j Considerations Relevant Regulations Regulatory Experience and Regulatory Guides and Position l
- 1 A.8 Industrial, Military, and 2 Transportation Fac8ities 3 Accidents at present or f[ 4 projected nearby industrial, The proposed Section 100.21,
*Nonseismic Siting Criteria.*
Potentially hazardous facilities and i 5 military, and transportation activities within 8 km (5 mi), and
- . 6 facilities may affect the safety major airports within 16 km (10 7 10 CFR Part 50, Appendix A, mi), of a pioposed site should be of the nuclear power station. " General Design Cnteria for 8 idenafied, if a preliminary Nuclear Power Plants," evaluation of potential accidents
- 9 Criterion 4, " Environmental 10 at these facilities indicates that 11 and Dynamic Effects Design the potential hazards from shock Bases."
waves and missiles approach or 12 ~ exceed those of the design basis
- 13 Regulatory Guide 1.70, tomado for the region, or potential j 14 " Standard Format and Content hazards such as flammable vapor j 15 of Safety Analysis Reports for clouds, toxic chemicals, or
- 16 Nuclear Power Plants," incendiary fragments exist, the i 17 Section 2.2' (lists types of suitability of the site should be facilities and potential determined by detailed evaluation 18
{ accidents). of the potential hazard. i 19 l 20 Regulatory Guide 1.78, The acceptability of a site 21
- Assumptions for Evaluating depends upon establishing that (1) j the Habitability of a Nuclear
- 22 an accident at a nearby facility or
- 23 Power Plant Control Room route will not result in radiological ~.
i During a Postulated Hazardous consequences that exceed the 24 25 Chemical Release."' dose guideline set forth in the ( j 26 proposed Section 50.34, or (2) i 27 the accident is sufficiently unlikely i 28 to occur that it poses no undue 2D risk, or (3) the nuclear power 30 station can be designed so its 31 safety will not be affected by the accident. 32 33 The identification of design basis 34 events resulting from the presence of nearby hazardous 36 materials or activities in the vicinity of a nuclear power station 38 is acceptable if the design basis 39 events include each postulated 40 type of accident for which a 41 realistic estimate of the probability 42 of occurrence of a potential dose 43 in excess of that set forth in the 44 proposed Section 50.34 guideline 45 exceeds approximately 104 per year. A-10
. . l f
1 APPENDIX B 1 ! 2 ENVIRONMENTAL CONSIDERATIONS FOR ASSESSING 3 SITES FOR NUCLEAR POWER STATIONS l 4 This appendix summarizes environmental considerations related to site characteristics that j 5 should be addressed in the early site selection process. The relatia importance of the different 6 factors to be considered varies with the region or State in which the potential sites are located, f 7 Site selection processes can be facilitated by establishing limits for various parameters based on 8 the best judgment of specialists knowledgeable of the region under consideration. For example, l 9 limits can be chosen for the fraction of water that can be diverted in certain situations wrthout 10 adversely affecting the local populations of important species. Although simplistic because 11 important factors such as the distribution of important species in the water body are not taken into i i 12 account, such limits can be usefulin a screening process for site selection. 4 13 i 4 4 1 4 e i G I B-1 I
- - . - - . .-- - - --.- .- _.- - - . - . - - - _ - - . - . - . - . - . ~ -
~ .
Considerations Parameters Regulatory Position 1 B.1' Preservation of important 2 Habitats 3 Important habitats are those The proportion of an 4 that are essential to in general, a detailed jusofication important habitat that would should be provided when the 5 maintaining the reproductive 8 be destroyed or significantly destructen or significant capacity and vitality of 7 altered in relation to the total alteration of more than a few important species populations habitat within the region in 8 (defined in the Discussion percent of important habitat types which the proposed site is to is proposed. 9 section of this guide under 10 be located is a useful Ecological Systems and Biota) parameter for estimateg 11 or the harvestable crop of The reproductive capacity of potentialimpacts of the populations of important species 12 economically or recreationally 13 construction or operation of a and the harvestable crop of important4pecies. Such nuclear power staten. The 14 habitats include breeding economically or recreasonally value of the proportion varies important populations must be 15 areas (e.g., nesting and 16 among species and among maintained unless jusofication for spawning areast, nursery, habitats. The region consid-17 feeding, resting, and wintering proposed or probable changes can ered in determining pro- be provided. 18 areas or other areas of portions is the normal 19 seasonally high concentrations geographic range of the 20 of individuals of important specific population in 21 species, question. 22 The construction and opera- If endangered or threatened 23 tion of nuclear power stations species occur at a site, the 24 (including new transmission potential effects of the 25 lines and access corridors con- construction and operation of 26 structed in conjunction with a nuclear power station should 27 the station) can result in the , 28 be evaluated relative to the destruction or alteration of potential impact on the local 29 habitats of important species population and the total 30 leading to changes in the estimated population over the 31 abundance of a species or in entire range of species. 32 the species composition of a 33 community. See also Chapter 2 of 34 Regulatory Guide 4.2, 35
" Preparation of Environmental 36 Reports for Nuclear Power 37 Stations.
38 39
' Copies are available for inspection or copying for a fee from the NRC Public Document Room at 2120 40 41 L Street NW., Washingtoni DC; the PDR's mailing address is Mail Stop I.1.-6, Washington, DC 20555; telephone (2021634-3273; fax (202)634 3343. Copies may be purchased at current rates frm the 42 U.S. Govemment Printing Office, P.O. Box 37082, Washington, DC 20402 9328 (telephone (202)512-43 44 2249); or from the National Technical information Service by wnting NTIS at 5285 Port Royal Road, Springfield, VA 22161. '
B-2
! . a I
Parameters Regulatory Position Considerations i a (^) L - 1 B.2 Migratory Routes of . 2 Important Species Seasonal or daily migrations The vndth or cross sectional Narrow reaches of water bodies ( 3 are essential to maintaining area of a water body at a should be avoided as sites for 4 the reproductive capacity of proposed site relative to the locating intake or discharge i 5 some important species general vndth or cross- structures. 6 populations. sectional area in the portion of < 7 A zone of passage that will permit the water used by migrating ' species should be estwnsted. normal movement of irpportant 8 Disruption of migratory species populations and 9 patterns can result from partial or complete blockage Suggested minimum zones of maintenance of the harvestable 10 of migratory routes by passage range from 1/3 to crop of economically important 11 l structures, discharge plumes, 3/4 of the width or cross- populations should be provided. 1 12 13 environmental alterations, or sectional areas of narrow 14 human activities (e.g., trans- water bodies.2d 15 portation or transmission 16 corridor clearing and site Some species migrate in 17 preparation). central, deeper areas while 18 others use marginal, shallow 19 areas. Rivers, streams, and 20 estuaries are seldom 21 homogeneous in their lateral dimension with respect to ( 22 23 24 depth, current velocity, and habitat type. Thus, the use of width or cross-sectional area 25 26 criteria for determining . 27 adequate zones of passage 28 should be combined with a 29 knowledge of important 30 species and their migratory 31 requirements. 32 33 8 Water Quality Criteria. National Academy of Science's - National Academy of i 34 Engineering, Washington, DC,1972. 35 8 Handbook of Environmental Control. Volume Ill: Water Sunolv and Treatment. R.G. 36 Bond and C.P. Straub (Editors), CRS Press, Cleveland, Ohio,1973. O B-3
l Considerations Parameters l Regulatory Position i 1 B.3 Entrainment and 2 Impingement of Aquatic 3 Organisms 4 Plankton, including eggs, The depth of the water body The site should have charac-5 larvae, and Juvenile fish, can at the point of intake relative teristics that allow placement of 6 be killed or injured by to the general depth of the 7 entrainment through power intake structures where the water body in the vicinity of relative abundance of important i 8 station cooling systems or in the site. species is small and where low 9 discharge plumes. approach velocities can be The proportion of water attained. (Deep regions are 10 The reproductive capacity of 11 withdrawn relative to the net generally less productive than important species' populations 12 new available water at the dM5w areas. It is not implied may be impaired by lethal 13 site is an indirect measure of that a' nthic intakes are stresses or by sublethat the destruction of plankton, 14 necessary.) stresses that affect reproduc- which in turn is indicative of 15 tion of individuals or result in possible effects on important habitats (see 8.1 of this 16 increased predation on the populations of important Appendix B) should be avoided as 17 a'fected species population. species. It has been locations for intake structures. suggested that the fraction of 18 Fish and other aquatic 19 available new water that can organisms can be killed or be diverted is in the range of 20 injured by impingement on 21 10% to 20% of flow.5* cooling water intake screens'. 22 or by entrainment in discharge The simplistic parameter 23 plumes. (proportion of water with-24 25 drawal) is suitable for use in a screening process or site 26 selection. However, other 27 28 factors such as distribution of important species should be 29 30 considered and in all cases the advice of experts on the local 31 ; 32 fisheries should be consulted ' to ensure that proposed l 33 withdrawals will not be 34 excessive. l 35 36
- Approach velocity and screen-face velocity are design criteria that may affect the impingement of 37 larger organisms, principally fish, on intake screens. Acceptable approach and screen-face velocities 38 are based on swimming speeds of fish, which will vary with the species, site, and season.
39
- The Water's Edoe: Critical Problems of the Coastal Zone. B.H. Ketchum (Editor), MIT Press, 40 Cambridge, Mass.,1972.
41
- Engineering for Resolution of the Energy-Environment Dilemma, National Academy of Engineering, 42 Washington, DC,1972. ,
l l l i B-4 , I i l
Regulatory Position Considerations Parameters A 1 B.4 Entrapment of Aquatic 2 Organisms Site characteristics that will Sites where the construction of 3 Cooling water intake and intake or discharge canals would discharge system features, accommodate design features 4 be necessary should be avoided that mitigate or prevent 5 . such as canals and thermal unless the site and important plumes, can attract and entrap entrapment. 6 species characteristics are such 7 organisms, principally fish. that entry of important species to 8 The resulting concentration of the canal can be prevented or 9 important fish species near limited by screening. 10 the station site can result in 11 higher mortalities from station-l
, 12 related causes, such as im-13 pingement, cold shock, or gas 14 bubble disease, than would ! 15 otherwise occur.
16 Entrapment can also interrupt 17 normal migratory patterns. 18 B.5 Water Ot,ality Effluents discharged from Applicable EPA-approved Pursuant to Section 401(a)(1) of 19 the FWPCA, certification from the State water quality standards. O
! ) 20 nuclear power plants are governed under the authority State that any discharge will 21 comply with applicable effluent 22 of the Federal Water Pollution For states without EPA-approved water quality limitations and other water pollo-23 Control Act (FWPCA)-(PL 92- tion control requirements is standards, the water quality 24 500). necessary before the NRC can criteria listed in Water Quality 25 issue a construction permit un! ass Criteria,19728 will be used 26 the requirement is waived by the g for evaluation.
27 State or the State fails to act 28 within a reasonable length of 29 time. 30 issuance of a permit pursuant to 31 Section 402 of the Act is not a 32 prerequisite to an NRC license or i 33 permit. 34 Where station construction or 35 operation has the potential to 36 degrade water quality to the pos-37 sible detriment of other users, 38 more detailed analyses and 39 evaluation of water quality may 40 be necessary. 41 O 1 B-5
Considsrations Parameters Regulatory Position 1 B.6 Water Availability I 2 The consumptive use of water Applicable Federal, State, and 3 for cooling may be restricted Water use and consumption must local statutory requirements. comply with statutory 4 by statute, may be 5 requirements and be compatible inconsistent with water use Compatability with water use 6 planning, or may lead to an with water use plans of cognizant plan of cognizant water water resources planning 7 unacceptable impact to the resource planning agency, agencies. 8 water resource. 9 10 in the absence of a water use Consumptive use should be plan, the effect on other restricted such that the supply of 11 water users is evaluated, other users is not impaired and ( 12 . considering flow or volume 13 that applicable surface water 14 reduction and the resultant quality standards could be met, ability of all users to obtain assuming normal station 15 adequate supply and to meet operational discharges and 16 applicable water quality extreme low flow conditions 17 i standards (see B.5, Water defined by generally accepted l 18 Quality, of this appendix). engineering practices. 19 20 For multipurpose impounded lakes 21 and reservoirs, consumptive use 22 should be restricted such that the 23 magnitude and frequency of 24 drawdown will not result in 25 unacceptable damage to impor-26 tant habitats (see B.1, Preser-27 vation of important Habitats, of this appendix) or be inconsistent 28 29 with the management goals for the water body. 30 B.7 Established Public 31 Amenity Areas 32 Areas dedicated by Federal, Proximity to public amenity Siting in the vicinity of designated 33 State, or local govemments to area. Viewability (see B.10, 34 public amenity areas will generally scenic, recreational, or cultural Visual Amenities of this require extensive evaluation and 35 purposes are generally prohi- Appendix). justification. 36 bited areas for siting power 37 stations. The evaluation of the suitability of sites in the vicinity of public 38 Siting nuclear power stations amenity areas is dependent on 39 in the vicinity of established consideration of a specific plaat 40 public amenity areas could design and station layout in 41 result in the loss or deteriora- relation to potentialimpacts on 42 tion of important public the public amenity area. 43 amenities. B-6
- - .-- - .. .- - - - - - - .-- .-=.
C,
- l 1
1 i Parameters Regulatory Position l ! Considerations i [ i 1 8.8 Prospective Designated 2 Amenity Areas Comparison of possibk Public amenity areas that are 3 Areas containing important distinctive, unique, or rare in a resources for scenic, recros- amenity areas in number and 4 region should be avoided as sites 5 teonal, or cultural use may not extent with other similar areas available on a local, regional, for nuclear power statums. l 6 currently be designated as l such by public agencies but or national basis, as 7 ; 8 may involve a not loss to the appropriate. 9 public if converted to power 10 generation. These areas may I 11 include. locally rare land types, 12 such as sand dunes, wet-13 lands, or coastal cliffs. l 14 8.9 Public Planning l Officially adopted land use Land use plans adopted by 15 Land use for a nuclear power plans. Federal, State, regional, or local j 16 station should be compatible government entities must be 17 with established land use or examined, and any conflict 18 zoning plans of governmental between these plans and use of a
. 19 entities. proposed site must be resolved by 20 consultation with the appropnate 21 governmental entity.
22 23 8.10 Visual Amenities The presence of power stdon The solid angle subtended by The visualintrusion of nuclear
. 24 power station structures as 25 structures may introduce station structures at critical viewing points. viewed from nearby residential, 26 adverse visualimpacts to resi-recreational, scenic, or cultural 27 dential, recreational, scenic, or areas should be controlled by 28 cultural areas or other areas selecting sites where existing 29 with significant dependence topography and forests can be 30 on desirable viewing utilized for screening statum 31 characteristics.
structures from those areas in l 32 which visualimpacts would 33 otherwise be unacceptable. 34 I 35 8.11 Local Fogging and icing The hazards on transportation 36 Water and water vapor increase in number of hours of fogging or icing caused by routes from fog or ice that result 37 released to the atmosphere operation of the station. from station operation should be 38 from recirculating cooling systems can lead to ground evaluated. The evaluation should 39 include estimates of frequency of 40 fog and ice, resulting in occurrence of station-induced fog-41 transportation hazards and ging and icing and their impact on damage to electric G424344 transmission systems. transportation, electrical trans-mission, and other activities and { i functions. 45 B-7 s ..
Considerations Parameters Regulatory Position 1 B.12 Cooling Tower Drift 4 2 Concentrations of chemicals, The percent drift loss from The potential loss of important 3 dissolved solids, and recirculating condenser terrestrial species and other 4 cuspended solids in cooling cooling water, particle size resources should be considered. 5 tower drift could affect ter- distribution, salt deposition 6 restrial biota and result in rate, local atmospheric condi-7 unacceptable damage to tions, and loss of sensitive 8 vegetation and other terrestrial biota affected by 9 resources. salt deposition from cooling 1 10 l tower drift. 11 B.13 Cooling Tower Plume 12 Lengths 13 Natural draft cooling towers The number of hours per year The visibility of cooling tower 14 produce cloud-like plumes that the plume is visible as a plumes as a function of direction l 15 vary in size and altitude function of direction and dis- and distance from cooling towers 16 depending on the atmospheric tance from the cooling should be considered. The evalu-17 conditions. The plumes are towers. ation should include estimates of 18 usually a few miles in length frequency of occurrence for 19 before becoming dissipated, plumes as well as potential 20 although plume lengths of 20 hazards to aviation in the vicinity 21 to 30 miles have been of commercial and military 22 reported from cooling towers, airports. 23 Visible plumes emitted from 24 cooling towers could cause a 25 hazard to commercial and 26 military aviation in the vicinity 27 of commercial and military 28 airports. The plumes 29 themselves or their shadows 30 could have aesthetic impacts. 31 B.14 Plume Interaction 32 Water vapor from cooling The degree to which impacts The hazards to public health, 33 tower plumes may interact may occur will vary depending structures, and other resources 34 with industrial emissions from on the distance between the from potential plume interaction 35 nearby facilities to form nuclear and fossil-fueled sites, between cooling tower plumes 36 noxious or toxic substances the hours per year of plume and plumes from fossil-fueled 37 that could cause adverse interaction, the type and sites and industrial emissions from 38 public health impacts, or concentration of chemical nearby facilities should be 39 result in unacceptable levels reaction products, the area of considered. 40 of damage to biota, chemical fallout, and the local 41 structures, and other atmospheric conditions. 42 resources. G B-8
\
I n Considerations Parameters Regulatory Position [ 1
- (
1 8.15 Noise 2 Undesirable noise levels at 3 Applicable Federal, State, and Noise levels at proposed sites nuclear power stations could local noise regulations. } 4 occur during both the must comply with statutory 1 5 construction and operation requirements. 6 phases and have unacceptable
- 7 impacts near the plant.
8 8.16 Economic impact of 9 Preemptive Land Use 10 Nuclear power stations can 11 preempt large areas, The level of local economic if a preliminary evaluation of net 12 dislocation, such as loss of especially when large cooling local economic impact of the use 13 income, jobs, and production, of productive land for a nuclear lakes are constructed. The caused by preemptive use of 14 1:nd requirement is likely to be power station indicates a potential 15 productive land and its effect an important issue when a . for large economic dislocation, the 16 on meeting foreseeable ; proposed site is on productive NRC staff will require a detailed 17 national demands for agricul- evaluation of the potentialimpact land (e.g., agricultural land) ture products. 18 that is locally limited in avail- and justification for the use of the 19 ability and is important to the site based on a cost effectiveness 20 local economy, or which may comparison of alternative station g ) 21 be needed to meet foreseeable designs and site-station combina-V 22 national demands for agri- tions. To complete its evaluation, 23 cultural products. the staff will also need informa-24 tion on whether and to what 25 extent the land use affects 26 national requirements for agricul-27 tural products. B.17 Environmental Justice 28 A proposed site could have 29 Applicable Federal, State, and Areas that disproportionately inequitable impacts on 30 local statutory and regulatory affect minority or low-income minority and low-income requirements. 31 communities. populations should be avoided as sites for nuclear power stations. C L B-9
1 DRAFT REGULATORY ANALYSIS 2 A separate regulatory analysis was not prepared for this guide. The 3 draft regulatory analysis, " Proposed Revisions of 10 CFR Part 100 and 10 CFR 4 Part 50," was prepared for the proposed amendments, and it provides the 5 regulatory basis for this guide and examines the costs and benefits of the !, 6 rule as implemented by the guide. A copy of the draft regulatory analysis is j ,7 available for inspection and copying for a fee at the NRC Public Document j 8 Room, 2120 L Street NW., (Lower Level), Washington, DC, as Enclosure 2 to l 9 SECY 94-194. 1 l 1 3 1~ ] i 1 i i 2 i i O i RA-1
, Action: Morrison,RESL )
Cyr, v6t, y f#puog*g ' UNITED STATES
@34' i o NUCLEAR REGULATORY COMMISSION Cys: Taylor PPN '
5 W ASHINGTON, D.C. 20555
. ,1 Milhoan S 4 IN RESPONSE, PLEASE
"% ...+ /
- REFER TO: M960612 orrtCE OF THE SECRETARY yulY 7' 1ggg bWU Blaha Soffer AMurphy, RES MEMORANDUM TO: James M. Taylor " '
i Executive Director for Operations l g.g FROM: *UdI1n C. Hoyle, Secretary
SUBJECT:
)~
STAFF REQUIREMENTS - BRIEFING ON PART 100 FINAL RULE ON REACTOR SITE CRITERIA (SECY l 118), 3:00 P.M., WEDNESDAY, JUNE 12, 1996, i COMMISSIONERS' CONFERENCE ROOM, ONE WHITE FLINT NORTH, ROCKVILLE, MARYLAND (OPEN TO PUBLIC ATTENDANCE) l The Commission was briefed by the NRC staff on the Final Rule on i Reactor Siting Criteria revising the regulations on seismic and l geologic criteria and on radiological criteria,
- i. l I
The Commission requested the staff to determine whether the ! l proposed final rule would overrule or conflict with the Commission's decision in Seabrook that emergency. planning is not site disqualifying. 1
'(E90/OGC) (Secy Suspense: k/ h/) (9600091)
RES ' The Commission indicated that it looked forward to the work that the staff was doing on the pilot program for applying the new source term to operating plants and asked the staff to come back to the Commisson for guidance as it proceeds. The staff should be cautious that they do not proceed down a path of de facto exemptions'if rule changes are necessary to allow application of j the new source term to operating reactors on a generic basis. 4gpo.)_ (RES) (Secy suspense: 11/1/96) (9600092) cc: Chairman Jackson Commissioner Rogers l- Commissioner Dicus [ OGC OCA , OIG i Office Directors, Regions, ACRS, ACNW, ASLBP (via E-Mail) PDR - Advance 1 I ! DCS - P1-24 i 1 s I hh"}}